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PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationAPRIL 15-20, 2012AcknowledgementThe task of organizing a large international topic meeting such as PHYSOR requires a tremendous amount of effort froma dedicated team. On behalf of the PHYSOR 2012 Organizing Committee we thank the ANS Reactor Physics Divisionas the primary sponsor, the ANS Mathematics and Computations Division as co-sponsor, and particularly the ANS OakRidge/Knoxville Local Section in providing local support. Additionally, the University of Tennessee Student Section providessupport in the meeting preparation as well as support during the meeting. ANS National provided support in the review ofcontracts, advertising and other support.The primary attraction for this meeting is the strong technical program. The international reactor physics community is wellknown for strongly supporting this series of meetings with outstanding technical papers, presentations, and posters. Wewould like to thank all of the authors and presenters for their technical contributions to the meeting. The technical programis the primary responsibility of the Technical Program Chair and for a meeting of this size is a significant task and is themost challenging aspect of organizing the meeting. We would like to provide our deepest thanks and gratitude to Ron Elliswho worked tirelessly through development of the call for papers, organizing the international technical program committee(TPC), coordinating paper reviews and final paper submissions. We also acknowledge the support provided by the technicalprogram co-chairs and the TPC members.In addition to the technical program, there are numerous other arrangements that must be made. We appreciate the excellentwork by the Katherin Goluoglu for meeting arrangements, Trent Primm as finance chair, Josh Peterson for specialevents, Germina Ilas for publications, Brian Ade for technical tours, Chris Perfetti for transportation, Dan Ilas for workshops,Lane Carasik for student support, and the outstanding effort of our sponsorship chair Chris Robinson, in addition to severalothers who have provided their support to this meeting. We were fortunate to have such dedicated and professional peopleinvolved in all of these important activities.The website, paper review, program and other support was provided by Hanna Shapira from Tech-info Comprehensive Solutions(TICS). Hanna worked as an integral part of the organizing team and her excellent support and experience is a bigfactor in the successful outcome of this meeting.Finally we provide our sincere thanks to the significant contributions by the meeting’s financial sponsors. These sponsorshipcontributions provide resources to make the meeting more enjoyable through support of breaks, meals, and otheractivities. The contributions are also used to allow us to maintain a low registration fee for the student participants in themeeting. Please refer to this program and our website for a list of all the sponsors.Best RegardsJess Gehin & Ivan MaldonadoCo-General Chairs2

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationAPRIL 15-20, 2012Platinum Sponsors3

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationAPRIL 15-20, 2012Organizing CommitteeHonorary Chairs Lee Dodds UTKKord Smith MITPaul Turinsky NCSUGeneral Co-Chairs Jess Gehin ORNLIvan Maldonado UTKTPC Ron Ellis ORNLCo-chairs Rakesh Chawla PSISandra Dulla Pol. di TorinoDeokjung Lee UNISTJim Lemons TVABojan Petrovic GTPiero Ravetto Pol. di TorinoFinance Chair Trent Primm PCLLCCorp Sponsorships Chris Robinson Y12Int’l Relationship Jim Gulliford OECD/NEAGary Dyck IAEAExhibits/Arrangements Katherin Goluoglu NSAWeb Applications Hanna Shapira TICSGuest ProgramWendy EllisPublicity Ugur Mertyurek ORNLTechnical Tours Brian Ade ORNLPublications Germina Ilas ORNLSpecial Events Josh Peterson ORNLWorkshops Dan Ilas ORNLTransportation Chris Perfetti ORNLStudent Chair Lane Carasik UTK ANSStudentChapterDodds Smith Turinsky Gehin MaldonadoEllis R Primm Robinson Goluoglu Ilas GIlas D Shapira Mertyurek Ellis W PetersonGulliford Ade Perfetti Carasik6

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationAPRIL 15-20, 2012Technical Program CommitteeTechnical Program Chairs Ron Ellis ORNL USACo-chairs Rakesh Chawla PSI SwitzerlandSandra Dulla Politecnico di Torino ItalyDeokjung Lee UNIST Rep. of KoreaJim Lemons TVA USABojan Petrovic Georgia Tech USAPiero Ravetto Politecnico di Torino ItalyTechnical Program Committee Members Track LeadsDr. Ronald J. EllisDr. Harold J. SmithProf. Andrew WorrallDr. Patrick BlaiseDr. Baocheng ZhangDr. David DiamondDr. Forrest B. BrownMr. Ian C. GauldDr. Eleodor M. NichitaMrs. Joy D. ForsterDr. Yi-Kang LeeDr. Kostadin N. IvanovDr. Richard SanchezDr. David P. GriesheimerMr. Peter GrimmMr. William J. MartinBrian J. AdeDr. Joshua L. PetersonDr. Andrea Borio Di TiglioleDr. Liangzhi CaoDr. J. Eduard HoogenboomDr. Luiz C. LealMr. Trent PrimmDr. Lei ZhuMr. Mick MastilovicDr. Sandro PelloniMr. Jeremy A. RobertsMr. Paul K. RomanoDr. Tomohiro EndoDr. Thomas M. SuttonDr. Irina PopovaDr. Winfried ZwermannDr. Blair P. BromleyDr. Tatiana IvanovaDr. Brian C. KiedrowskiDr. David W. NiggDr. Shinya KosakaDr. Vefa KucukboyaciDr. Farzad RahnemaDr. Daniel F. GillDr. Bo ShiDr. Toshikazu TakedaDr. Mario CartaDr. Dan IlasDr. Germina IlasDr. Flavio D. GiustDr. Wei ShenMr. Hakim FerroukhiDr. Nam Zin ChoDr. Mark DeinertDr. Josh JarrellDr. Temitope A. TaiwoDr. Jesse CheathamProf. Dr. Hans-Dieter BergerMr. Michael S. DefilippisDr. George F. FlanaganDr. Michaele C. Brady RaapDr. Hideki MatsumotoDr. Takanori KitadaMr. Martin A. ZimmermannDr. Gerald RimpaultDr. Gert Van Den EyndeDr. Paolo FerroniDr. Charles F. WeberDr. Glenn E. SjodenDr. Alexander VasilievProf. Christophe R. DemaziereDr. Peter L. AngeloDr. Go ChibaMr. Erik KolstadDr. Sandra DullaMr. Yuichiro BanMr. Noriyuki TakemotoDr. Alireza HaghighatDr. Ron DaganMr. Frederik ReitsmaDr. Ken NakajimaDr. Mark L. WilliamsKimberly SissomDr. Annalisa ManeraDr. Dimitrios CokinosDr. Richard D. McknightDr. Brian R. NeaseProf. Piero RavettoDr. Wolfgang WiesenackDr. Frédéric LaugierDr. Toshihiko KawanoDr. Jun-Ichi KatakuraDr. Hiroyuki OigawaMr. Tsuyoshi YamaneDr. Fabian Eduardo. JatuffDr. Maria N. AvramovaDr. Sara BortotDr. Taraknath WoddiMr. Henry PueschelDr. Nicolay SultanovDr. Alan H. TkaczykProf. Pedro VazProf. Francesco D’auriaDr. Emily R . WoltersDr. Alessandro AlembertiDr. Enrico SartoriMr. Nigel T. (Jim). GullifordDr. Timothy E. ValentineIgor ZmijarevicDr. Keisuke KobayashiMr. Massimo SalvatoresMr. Marcel BouffierDr. Kevin T. ClarnoDr. Deokjung LeeDr. Massimiliano RosaMargaret EmmettDr. Akio YamamotoDr. Won Sik YangDr. Guillermo Ivan. MaldonadoDr. Abdelhamid DokhaneMohamed OuisloumenDr. Gregory PerretDr. Rene Van GeemertDr. Cheolho PyeonDr. Mike WestfallDr. Jess GehinDr. Gary DyckDr. Fausto FranceschiniDr. Bojan PetrovicDr. Pavel TsvetkovDr. Tom Downar1 Core Analysis MethodsTom DownarNam Zin Cho2 Deterministic Transport TheoryCristian Rabiti USAFarzad Rahnema USAJ. E. Morel USAUSASouth Korea3 Monte Carlo Methods and DevelopmentsDavid Griesheimer USAJ.E. Hoogenboom NetherlandsThomas Sutton USA4 Reactor Concepts and DesignsPavel Tsvetkov USAHans Dieter Berger Germany5 Education in Reactor PhysicsBernadette KirkCheolho Pyeon6 Reactor Operation and SafetyAndy WorrallKen KozierUSAJapanUKCanada7 Fuel Cycle and Actinide ManagementMick Mastilovic USATemi TaiwoUSA8 Advanced Modeling and Simulation in ReactorPhysicsJohn Wagner USAWilliam Martin USA9 Research Reactors and Spallation SourcesCheolho Pyeon JapanIrina Popova USA10 Nuclear Criticality SafetyM. C. Brady-Raap USABrad Rearden USA11 Nuclear DataLuiz LealUSA12 Sensitivity and Uncertainty AnalysisHany Abdel-Khalik USATatiana Ivanova France13 Fuel, Material, Mechanical Analysis andBehaviorMike Defilippis USAPaolo Ferroni USA14 Reactor Transient and Safety AnalysisK. Ivanov USA15 Experimental Facilities and ExperimentsP.D. BlaiseFrance16 Radiation Applications and Nuclear SafeguardsSara PozziUSASpecial Session 1 on Reactor Physics ExperimentsIn honor of Nils Göran Sjöstrand, recipientof the 2011 Eugene Wigner’s PrizeImre PazsitSwedenPiero Ravetto ItalyDimitrios Cokinos USASpecial Session 2: Radiation Transport Methodsfor Whole Reactor Core AnalysisFarzad Rahnema USAThomas Evans USAAlireza Haghighat USA7

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationAPRIL 15-20, 2012Extra EventsMonday Night: Women’s Basketball Hall of Fame$40 per extra ticket - covered in regular registrationMonday Night at the Women’s Basketball Hall of Fame willbe the place to be. The Women’s Basketball Hall of Fameopened in June 1999 in Knoxville, Tennessee. It is the onlyfacility of its kind dedicated to all levels of women’s basketball.So whether you are looking for inspiration, education or justplain fun, the Women’s Basketball Hall of Fame is the placeto find it!The guest speaker will be Joan Cronan. Joan Cronan hasbeen the women’s athletic director at the University of Tennessee,Knoxville since 1983. She was named National Associationof Collegiate Women Athletics Administrators 2005athletic director of the year. Under her direction, the LadyVols program has captured the Southeastern Conference’swomen’s all-sport award from the New York Times for threeconsecutive years. She will be speaking about the past,present, and future of women’s sports at UT-Knoxville. It willbe a great honor to hear her speak.Enjoy a delicious meal catered by the world’s largest fine diningcompany, Ruth’s Chris Steak House. The meal will includepassed hors D’oeuvres such as mini sweet potato casseroleand seared Ahi tuna on cucumber. There will also be a beeftenderloin carving station for the meat eaters out there, andof course, dessert is a must.The basketball courts are located downstairs and will beavailable for some friendly competitions during the night.Who knows maybe the engineers at ORNL will be able tooutplay the engineers at INL and ANL. Only one way to findout is on the courts.The doors will open at 6:30 pm and the last call for drinkswill be at 9:30 pm. Joan Cronan will speak around 7:30. Thebuses will be on a rotating schedule so you may come andleave while you please.Tuesday Night: Three River Rambler$40 - limited to 125 peopleGuests aboard the excursion train will enjoy a two-hour tripalong the Tennessee River. Beginning the journey in DowntownKnoxville, the Rambler travels past historical sites to the“Three Rivers Trestle” where the French Broad and HolstonRivers join to form the Tennessee River.Along the route, the Three Rivers Rambler passes by beautifulfarmland, Knoxville’s first settlement area, and severalquarries that were mined to build our nation’s Capital. Theuniformed conductor and volunteer staff are well versed in thelocal history and welcome questions during the 11-mile trip.To go along with the old-fashioned feel of the train ride, thefood will be catered by Calhon’s Restaurant. Enjoy the HickorySmoked Pork BBQ and BBQ Roasted Chicken. Try somesides like smoky mountain baked beans and cole slaw, andof course, no old-fashioned train ride is quite right withoutassorted fresh baked cookies.If you get thirsty while on the train ride you can enjoy locallybrewed beer on tap such as the tallship IPA or the Appalachianpale ale. We will also have locally made wine for those whoenjoy a more refined way of quenching your thirst, all includedin the price of the ticket.The buses for the train will leave at 5:30 pm and the trainwill depart the station at 6:00 pm. You will be able to enjoythe sunset from the train as you head back to arrive home ataround 8:00 pm.For the train buffs out there the track is owned by the Knoxville& Holston River Railroad (KXHR) who purchased it from NorfolkSouthern. KXHR is responsible for maintaining the trackand bridges and operating all the trains. The Federal RailroadAdministration and Tennessee Department of Transportationinspect the track to see that it meets safety standards. TheKXHR is owned by Gulf & Ohio Railways, a Knoxville basedcompany with shortline railroads in Alabama, Tennessee,and North Carolina. Gulf & Ohio Railways is a family ownedbusiness that has been hauling freight since 1985.Thursday Night: Sunshpere$40 - limited to 200 peopleThursday night the party will begin at the top of Knoxville,TN within the Sunsphere. The Knoxville Sunsphere was thetheme structure for the 1982 World’s Fair. It represents thesun, source of energy, and reflected the energy theme ofthe fair. 24-karat gold gives the panes of glass the reflectivegold color.The food will be catered by a locally owned restaurant calledCalhoun’s. It will included a carving station, a pasta station,stuffed mushrooms, and much much more. The open bar willinclude locally brewed beer from the smoky mountain brewery.Locally grown wine will also be among the many selectionsof drinks to choose from.The music will be Mountain Soul. Mountain Soul’s eclectic arrangementsand honest songwriting have meshed to becomean exciting new sound in Americana music. Brothers Danieland Cory Kimbro have continued in the dynamic performancetradition they learned from their parents. Jonathan Manessbrings the crucial third part to their brother harmony whileincorporating a unique flatpicking guitar style, and Mike Seal’srefreshing resophonic and electric guitar stylings completeMountain Soul’s version of Appalachian roots music.This will be an amazing ending to a great conference. Thedoors will open at 6:00 pm and the last call for drinks will beat 9:00 pm.9

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationAPRIL 15-20, 2012Things to do in KnoxvilleDowntown KnoxvilleDogwood Arts Festivalhttp://www.dogwoodarts.com(see website for a listing of events)Knoxville Museum of Art1050 Worlds Fair Park Drive(865) 934-2034http://www.knoxart.orgBijou Theatre803 S. Gay Street(865) 523-2665http://www.knoxbijou.comArts & Culture Alliance & Emporium Center100 S. Gay Street(865) 523-7543http://www.knoxalliance.comMarket Squarehttp://knoxvillemarketsquare.comVarious eateries, shops, entertainment, etc.Knoxville AreaKnoxville Zoo3500 Knoxville Zoo Drive(865) 637-5331http://www.knoxville-zoo.orgWest Town Mall (West Knoxville)7600 Kingston Pike(865) 693-4731Hours: Mon-Sat 10am-9pm; Sun-12pm-6pmhttp://www.simon.comTurkey Creek Shopping (West Knoxville)Various shops, dining, cinemashttp://www.turkeycreekshopping.comAmerican Museum of Science and Energy300 S. Tulane AvenueOak Ridge, TN(865) 576-3200Hours: Mon-Sat 9am-5pm; Sun-1pm-5pmhttp://www.asme.orgGreat Smoky Mountains National ParkGatlinburg, TN(865) 436-1200 (Visitor Information)http://www.nps.gov/grsm/index.htmPigeon Forge Shopping & Attractionshttp://www.mypigeonforge.com/whattodo_shopping.aspxDowntown DiningCafé 44 Market Square(865) 544-1444American cuisineDowntown Grill & Brewery424 S. Gay Street(865) 633-8111http://www.downtown brewery.comAmerican cuisineThe Golden Shisha609 James Agee Street(865) 951-0536Mediterranean cuisineThe Tomato Head12 Market Square(865) 637-4067http://www.thetomatohead.comAmerican cuisineThe Melting Pot (dinner only)111 N. Central Street(865) 971-5400www.meltingpot.comFondueNama Sushi Bar135 S. Gay Street(865) 633-8539http://www.namasushibar.comJapanese cuisineMarket Square Kitchen1 Market Square(865) 546-4212http://www.marketsquarekit.netAmerican cuisineTrio Café13 Market Square(865) 246-2270http://www.trio-cafe.netAmerican cuisineSoccer Taco9 Market Square(865) 544-4471http://www.soccertaco.comMexican cuisine10

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationSunday April 15, 2012Technical WorkshopsWORKSHOP Location Time Coordinator1 SCALE Sensitivity and Uncertainty Analysis Rm. 301 A 01:30pm – 05:30pm Brad Rearden2 Reactor Physics Analysis Using SCALE/TRITON Rm. 301 C 08:00am – 12:00pm Matt Jessee3 Hybrid Radiation Transport Methods for Whole-Reactor-CoreAnalysis and Nuclear Safeguards ApplicationsRm. 301 B 08:00am – 12:00pm Farzad Rahnema and AlirezaHaghighat4 Scientific Computing for Nuclear Engineering Rm. 301 A 08:00am – 12:00pm The Hacker Within(Anthony Scopatz)5 Advanced Monte Carlo for Reactor Physics Core Analysis Rm. 301 B 01:30pm – 05:30pm Forrest Brown6 PARCS Workshop: Nuclear Reactor Simulation Rm. 301 C 01:30pm – 05:30pm Tom Downar7 Advanced Reactor Concepts Rm. 301 D 08:00am – 05:30pm David Holcomb, Hans Gougar,and Taek K. Kim8 Short Course on Uncertainty Quantification & Sensitivity Ballroom A 08:00am – 12:00pm Hany Abdel-KhalikAnalysis Methods9 Convergence Acceleration and Reactor Physics Rm. 301 E 08:00am – 12:00pm Barry Ganapol10 Application of Attila for Reactor Analyses Rm. 301 E 01:30pm – 05:30pm Transpire, Inc. (GregoryFailla)11

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationMonday, April 16, 2012 - 8:00 AM - Ballrooms E-GPlenary Speakers8:40 AMSteven P. NesbitDirector – Nuclear Policy and SupportDuke Energy“Nuclear Electricity Generation at the Crossroads”Steve Nesbit is responsible for developing company policy positions related to nuclear power and interactingwith industry and government groups on used fuel management and related issues. Nesbit joined DukePower’s nuclear generation department in 1982. Since then, he has held various roles in nuclear safetyanalysis and fuel management. Nesbit received Bachelor of Science and Master of Engineering degrees innuclear engineering from the University of Virginia. He is a registered professional engineer in North Carolina and South Carolina. Nesbitis also an adjunct faculty member at the University of North Carolina at Charlotte where he teaches a nuclear engineering course.9:05 AMDr. Phillip J. FinckChief Nuclear Research OfficerIdaho National Laboratory“Integrated Fuel & Reactor Multiphysics and Safety Assessment Tools: Validation and ExperimentalNeeds”Dr. Finck is an internationally recognized expert in advanced reactor and fuel cycle systems. He is noted forhis technical leadership in reactor design and analysis, code development and validation, nuclear data, andmore recently, in advanced fuel cycles. Prior to joining Idaho National Laboratory, Dr. Finck worked at ArgonneNational Laboratory, where he was the associate laboratory director for Applied Science and Technology, and at the French Atomic EnergyCommission, where he was the head of the Reactor Physics Laboratory. Dr. Finck received his doctorate in nuclear engineering atMIT in 1982, and earned a Master of Business Administration from the University of Chicago. Dr. Finck is a Fellow in the American NuclearSociety.9:30 AMDr. Thom MasonLaboratory DirectorOak Ridge National Laboratory“Welcome Remarks”Dr. Mason became ORNL Laboratory Director in 2007. He received a B.Sc. in physics from Dalhousie Universityin Halifax, Nova Scotia, in 1986 and a Ph.D. in physics from McMaster University in Hamilton, Ontario, in1990. He was a postdoctoral fellow at AT&T Bell Laboratories and a senior scientist at at Risø National Laboratoryin Denmark. He was assistant and associate professor in the Department of Physics at the University ofToronto. He became director of the Experimental Facilities Division of the Spallation Neutron Source in 1998 and served in that capacityuntil being named Associate Laboratory Director for the Spallation Neutron Source in 2001. In 2006, Dr. Mason was named AssociateLaboratory Director for Neutron Sciences. He was an Alfred P. Sloan Research Fellow from 1997 to 1999, and has been an Associate ofthe Quantum Materials Program of the Canadian Institute for Advanced Research since 1993. He was elected a Fellow of the AmericanAssociation for the Advancement of Science in 2001 and a Fellow of the American Physical Society in 2007.12

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationMonday, April 16, 2012 - 8:00 AM (continued) - Ballrooms E-GPlenary Speakers9:40 AMDr. Susan MartinProvost and Senior Vice ChancellorUniversity of Tennessee – Knoxville“Welcome Remarks”Dr. Susan Martin was named Provost and Senior Vice Chancellor for Academic Affairs in May 2009. Previously,she had served as senior vice provost. Dr. Martin was associate dean of the UT Knoxville College ofArts and Sciences from 2000 to 2004, and prior to that spent nine years as a department head. She joinedthe faculty of the Department of Classics at UT Knoxville in 1981. Dr. Martin received B.A. degrees in comparativeliterature (French and German, 1973) and classical philology (1976) from the University of California, Berkeley. She has M.A. andPh.D. degrees in classical studies from the University of Michigan (1981). She spent a junior year abroad at the Université de Bordeaux(1971-72), was a fellow of the American Academy in Rome (1980-81) and a Liberal Arts Fellow in Law and Classics at Harvard LawSchool (1988-89). She has served as chair of the Advisory Council to the Classical School Committee of the American Academy in Romeand as President of the Tennessee Foreign Language Teaching Association. She was the recipient of the Jacqueline Elliott Award forService in Higher Education from the TFLTA in 2003. She is past president of the Classical Association of the Middle West and South.9:50 AMDr. Kord SmithHonorary ChairKorea Electric Power Co. Professor of the Practice of Nuclear Science and EngineeringMassachusetts Institute of Technology“Education and Research Paths Toward Reactor Analysis and Computational Methods in HighPerformance Computing Environments.”Since earning his SM (1979) and Ph.D. (1980) at MIT, Kord Smith has been instrumental in the developmentof the world’s most widely used software for reactor physics modeling and simulation, and also succeeded inbusiness as a co-founder of Studsvik Scandpower, a supplier of software for reactor core design, analysis, and operations to a majority ofthe world’s nuclear utilities and reactor fuel vendors. More recently, he was named chief scientist for the Center for Exascale Simulation ofAdvanced Reactors (CESAR), an interdisciplinary public-private effort of the Department of Energy’s Office of Science aimed at creatingan entirely new generation of nuclear-related high performance computing hardware and software technology.13

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationMonday, April 16, 2012 - 10:45 AM - Ballrooms E-GPlenary Speakers10:45 AMDr. Kurt EdsingerDirector, MaterialsElectric Power Research Institute (EPRI)“Breakthrough Fuel Designs”Dr. Kurt Edsinger is a Director at the Electric Power Research Institute (EPRI) overseeing the materials programsin EPRI’s Nuclear Sector. Research in the materials area is focused on improving our understandingof, and our ability to manage, primary systems materials degradation. Prior to managing the materials area,Dr. Edsinger was the director of Nuclear Fuel and Chemistry. Before joining EPRI, Dr. Edsinger worked forGeneral Electric’s fuels group and managed the Materials Technology Group for GE’s fuel business, Global Nuclear Fuel. In that position,he led a group of scientists and engineers in resolving BWR fuel performance issues, developing new fuel products, and demonstratingfuel reliability margins. Dr. Edsinger holds a Bachelor of Science degree in chemical engineering from San Jose State University. Hereceived a Doctorate degree in chemical engineering from University of California, Santa Barbara.11:10 AMCheri CollinsNuclear Development General Manager External AlliancesSouthern Nuclear Operating Co., Inc.“An Update on the Vogtle 3 & 4 Project and AP1000 Implementation”Cheri began her career in 1978 as a summer intern at Alabama Power’s Clanton District office. After graduatingfrom UAB engineering school in 1981, she began a career at the Joseph M. Farley nuclear plant nearDothan, Alabama. She spent the next 24 years at Plant Farley, obtaining a senior reactor operator’s license in1987 and concluding her time at Farley as the Plant manager from 2007 – 2009. She now works for NuclearDevelopment and is assigned to Southern Company’s corporate communications where she is a chief spokesperson for the Vogtle 3 and4 Project.11:35 AMDr. Roger ReynoldsSenior Technology AdvisorTerraPower“An Overview of the TerraPower Traveling Wave Reactor Technology”Dr. Reynolds has been involved with TerraPower since 2008 providing strategic counsel to TerraPower leadership.He has more than 40 years of broad based experience in the nuclear industry. He began his careeras a professor of nuclear engineering at Mississippi State University (MSU) where he was responsible forcourses in reactor engineering, nuclear fuel cycle analyses and reactor physics. During his tenure at MSU,he assisted with the initial start-up of the Grand Gulf Nuclear Station, and in the 1970s he served as a visiting scientist at -what is todayIdaho National Laboratory- performing fast reactor physics experiments at the Zero Power Plutonium (Physics) Reactor. After leaving theuniversity he worked 20 years for AREVA and its predecessors (Advanced Nuclear Fuels, Siemens Power Corp, Framatome) in severalpositions of increasing responsibility. Dr. Reynolds retired from AREVA in December 2007 as the Chief Technology Officer in the US andthe Chief Engineer for the US nuclear fuel business.14

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationMonday April 16, 2012 - 1:30 PM - 301 A1A - Core Analysis MethodsSession Chair: Brian Elder (TVA); Volkan Seker (U. Mich.)1:30 PMFlux Stabilization in Neutron Problems with Fixed SourcesDaniele Tomatis and Aldo Dall’OssoAREVA NP, Tour AREVA, Paris La D´efense Cedex, FranceAlthough critical core calculations are the most common in design and safety analysis,fixed source calculations are needed for specific applications, e.g. to compute ex-coredetector response functions, to develop new methodologies for dilution and reloaderror accidents and more in general for all situations involving sub-critical shut-downstates. It is well known that the source problem becomes difficult to be solved with coreconfiguration close to criticality, i.e. with the multiplication factor approaching unity,for the occurrence of numerical ill-conditioning and very high number of iterations,possibly leading to failure in the ux convergence. In this work, the Wielandt eigenshifttechnique used in iterative methods of critical problems is developed for source problemstoo, in order to stabilize the solution. The mathematical basis and the proof of theconvergence are discussed. Compared to the existing methods, this technique allowsalso more control to avoid singular behavior at inner iterations. Numerical tests with a1D analytical benchmark are reported to prove the robustness of the technique.1:55 PMQuasi-Heterogeneous Efficient 3-D Discrete Ordinates CAN-DU Calculations Using ATTILAPreeti T (1), Rulko R (2)1) University of Ontario Institute of Technology, ON, Canada. 2) Canadian Nuclear Safety Commission,Ottawa, Ontario, CanadaIn this paper, 3-D quasi-heterogeneous large scale parallel Attila calculations of ageneric CANDU test problem consisting of 42 complete fuel channels and a perpendicularto fuel reactivity device are presented. The solution method is that of discreteordinates SN and the computational model is quasi-heterogeneous, i.e. fuel bundleis partially homogenized into five homogeneous rings consistently with the DRAGONcode model used by the industry for the incremental cross-section generation. In calculations,the HELIOS-generated 45 macroscopic cross-sections library was used.This approach to CANDU calculations has the following advantages: 1) it allowsdetailed bundle (and eventually channel) power calculations for each fuel ring in abundle, 2) it allows the exact reactivity device representation for its precise reactivityworth calculation, and 3) it eliminates the need for incremental cross-sections. Ourresults are compared to the reference Monte Carlo MCNP solution. In addition, theAttila SN method performance in CANDU calculations characterized by significantupscattering is discussed.2:20 PMThe Explicit Representation for the Angular Flux Solution inthe Simplified PN (SPN) TheoryYung-An Chao (1) and Akio Yamamoto (2)1) (Retired) Apartment 101, Building 2, 788 Hong Xu Road, Shanghai, China, 2) Nagoya University:Furo-cho, Chikusa-ku, Nagoya, JapanThe current SPn theory formulation, via either the asymptotic method or the variationalmethod, does not provide an explicit and calculable representation for the correspondingangular flux solution. It is therefore not possible to reconstruct from the SPn solutionthe corresponding angular flux solution, or to extract from a reference transportsolution the corresponding SPn solution. This makes it impossible to calculate thenecessary surface discontinuity factors to force consistency between the SPn solutionand the higher level transport solution. Without discontinuity factors, the superiorityof SPn over diffusion could be significantly degraded in practical applications. In thispaper we present a different SPn formulation that provides the explicit angular fluxsolution such that the physical picture for the SPn approximation is transparent andthe SPn discontinuity factors can be calculated.2:45 PMAn Asymptotic Homogenized Neutron Diffusion Approximation.I. TheoryTravis J. Trahan and Edward W. LarsenDepartment of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MIA monoenergetic, homogenized, anisotropic diffusion equation is derived asymptoticallyfor large, 3-D, multiplying systems with a periodic lattice structure. The primaryassumption is that the system is slightly perturbed from an infinite, periodic lattice, andthat the length scale of a lattice element is small relative to the total system size. Theperturbed flux is slightly buckled, and the leading order term is the product of a slowlyvarying amplitude component, and a rapidly varying periodic component. The amplitudefunction is the solution to the homogenized diffusion equation, while the periodiccomponent is the solution to the unperturbed, infinite system, and can be found usingany high-order transport method. The first order term acts as a correction term, andmakes it possible to obtain a zero flux extrapolation distance for the diffusion equationby applying the Marshak boundary condition.3:10 PMAn Asymptotic Homogenized Neutron Diffusion Approximation.II. Numerical ComparisonsTravis J. Trahan and Edward W. LarsenDepartment of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MIIn a companion paper, a monoenergetic, homogenized, anisotropic diffusion equationis derived asymptotically for large, 3-D, multiplying systems with a periodic latticestructure [1]. In the present paper, this approximation is briefly compared to severalother well known diffusion approximations. Although the derivation is different, the asymptoticdiffusion approximation matches that proposed by Deniz and Gelbard, andis closely related to those proposed by Benoist. The focus of this paper, however, isa numerical comparison of the various methods for simple reactor analysis problemsin 1-D. The comparisons show that the asymptotic diffusion approximation provides amore accurate estimate of the eigenvalue than the Benoist diffusion approximations.However, the Benoist diffusion approximations and the asymptotic diffusion approximationprovide very similar estimates of the neutron flux. The asymptotic method andthe Benoist methods both outperform the standard homogenized diffusion approximation,with flux weighted cross sections.15

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationMonday April 16, 2012 - 1:30 PM - 301 B2A - Deterministic Transport TheorySession Chair: Steve Bowman (ORNL); Yunlin Xu (ANL)1:30 PMUsing Laguerre Polynomials to Compute the Matrix Exponentialin Burnup CalculationsDing She, Ang Zhu And Kan WangDepartment of Engineering Physics, Tsinghua University, Beijing, ChinaAn essential part of burnup analysis is to solve the burnup equations. The burnupequations can be regarded as a first-order linear system and solved by means ofmatrix exponential methods. Because of its large spectrum, it is difficult to computethe exponential of the burnup matrix. Conventional methods of computing the matrixexponential, such as the truncated Taylor expansion and the Pade approximation, arenot applicable to burnup calculations. Recently the Chebyshev Rational ApproximationMethod (CRAM) has been applied to solve burnup matrix exponential and shownto be robust and accurate. However, the main defect of CRAM is that its coefficientsare not easy to obtain. In this paper, an orthogonal polynomial expansion method,called Laguerre Polynomial Approximation Method (LPAM), is proposed to computethe matrix exponential in burnup calculations. The polynomial sequence of LPAM canbe easily computed in any order and thus LPAM is quite convenient to be utilized intoburnup codes. Two typical test cases with the decay and cross-section data takenfrom the standard ORIGEN 2.1 libraries are calculated for validation, against the referenceresults provided by CRAM of 14 order. Numerical results show that, LPAM issufficiently accurate for burnup calculations. The influences of the parameters on theconvergence of LPAM are also discussed.1:55 PMComparison of Direct and Quasi-Static Methods for NeutronKinetic Calculations with the EDF R&D Cocagne CodeE. Girardi and P. Guérin(1), S. Dulla, M. Nervo and P. Ravetto (2)1) Electricité de France – R&D, Clamart, France. 2) Dipartimento di Energetica, Politecnico di Torino,Torino, ItalyQuasi-Static (QS) methods are quite popular in the reactor physics community andthey exhibit two main advantages. First, these methods overcome both the limits ofthe Point Kinetic (PK) approach and the issues of the computational effort related tothe direct discretization of the time-dependent neutron transport equation. Second,QS methods can be implemented in such a way that they can be easily coupled tovery different external spatial solvers. In this paper, the results of the coupling betweenthe QS methods developed by Politecnico di Torino and the EDF R&D core code CO-CAGNE are presented. The goal of these activities is to evaluate the performances ofQS methods (in term of computational cost and precision) with respect to the directkinetic solver (e.g. scheme) already available in COCAGNE. Additionally, they allowto perform an extensive cross-validation of different kinetic models (QS and directmethods).2:20 PMMOCUM: A Two-Dimensional Method of Characteristics CodeBased on Unstructured Meshing for General GeometriesXue Yang and Nader SatvatSchool of Nuclear Engineering, Purdue University, West Lafayette, INA transport theory code MOCUM based on the Method of Characteristics (MOC) asthe flux solver with an advanced general geometry processor is developed for two-dimensionallattice and full core neutronics modeling. The core structure is representedby Constructive Solid Geometry (CSG) that uses Boolean operations to build complexgeometries from simple polygons. Arbitrary-precision arithmetic is also used in the processof building CSG objects to eliminate the round-off error from the commonly useddouble precision numbers. Then, the constructed core frame will be decomposed andrefined into a conforming Delaunay triangulation to ensure the quality of the meshes.The MOC solver kernel is fully paralleled using OpenMP and the developed numericalcode is validated by several benchmarks representing various core geometries. Forthe cases modeled, the maximum percentage error for multiplication factor and the pinpower compared to reference values are 0.1% and 0.7% respectively.2:45 PMApplication of the Discrete Generalized Multigroup Method toUltra-Fine Energy Mesh in Infinite Medium CalculationsNathan A. Gibson and Benoit ForgetDepartment of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge,MAThe Discrete Generalized Multigroup (DGM) method uses discrete Legendre orthogonalpolynomials to expand the energy dependence of the multigroup neutron transportequation. This allows a solution on a fine energy mesh to be approximated for a costcomparable to a solution on a coarse energy mesh. The DGM method is applied toan ultra-fine energy mesh (14,767 groups) to avoid using self-shielding methodologieswithout introducing the cost usually associated with such energy discretization. Resultsshow DGM to converge to the reference ultra-fine solution after a small numberof recondensation steps for multiple infinite medium compositions.3:10 PMDevelopment of a Coupling Code for PWR Reactor Cavity RadiationStreaming CalculationZheng Zheng, Hongchun Wu, Liangzhi Cao, Youqi Zheng, Hongbo Zhang,Mengqi WangNECP laboratory, School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an Shaanxi,China.PWR reactor cavity radiation streaming is important for the safe of the personnel andequipment, thus calculation has to be performed to evaluate the neutron flux distributionaround the reactor. For this calculation, the deterministic codes have difficultiesin fine geometrical modeling and need huge computer resource; and the Monte Carlocodes require very long sampling time to obtain results with acceptable precision.Therefore, a coupling method has been developed to eliminate the two problems mentionedabove in each code. In this study, we develop a coupling code named DORT-2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP isused to produce a combined surface source containing top, bottom and side surfacesimultaneously. Because SDEF card is unsuitable for the combined surface source,we modify the SOURCE subroutine of MCNP and compile MCNP for this application.Numerical results demonstrate the correctness of the coupling code DORT2MCNPand show reasonable agreement between the coupling method and the other twocodes (DORT and MCNP).16

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationMonday April 16, 2012 - 1:30 PM - 301 C4A - Reactor Concepts & DesignsSession Chair: Kim Sissom (TVA)1:30 PMPreliminary safety calculations to improve the design of MoltenSalt Fast ReactorM. Brovchenko, D. Heuer, E. Merle-Lucotte, M. Allibert, N.Capellan, V.Ghetta, A. LaureauLPSC, UJF, CNRS/IN2P3, Grenoble INP, Grenoble Cedex,FranceMolten salt reactors are liquid fuel reactors so that they are flexible in operation butvery different in the safety approach from solid fuel reactors. This study bears on thespecific concept named Molten Salt Fast Reactor (MSFR). Since this new nucleartechnology is in development, safety is an essential point to be considered all alongthe R&D studies. This paper presents the first step of the safety approach: the systematicdescription of the MSFR, limited here to the main systems surrounding the core.This systematic description is the basis on which we will be able to devise accidentalscenarios. Thanks to the negative reactivity feedback coefficient, most accidental scenarioslead to reactor shut down. Because of the decay heat generated in the fuel salt,it must be cooled. After the description of the tools developed to calculate the residualheat, the different contributions are discussed in this study. The decay heat of fissionproducts in the MSFR is evaluated to be low (3% of nominal power), mainly due tothe reprocessing that transfers the fission products to the gas reprocessing unit. Asa result, the contribution of the actinides is significant (0.5% of nominal power). Theunprotected loss of heat sink transients are studied in this paper. It appears that slowtransients are favorable (> 1min) to minimize the temperature increase of the fuel salt.This work will be the basis of further safety studies as well as an essential parameterfor the design of the draining system.1:55 PMEvaluation for 4S Core Nuclear Design Method Through Integrationof Benchmark DataA. Nagata, Y Tsuboi (1), Y. Moriki (2), M. Kawashima (3)1) Advanced System Design & Engineering Dept., Isogo Nuclear Engineering Center, Toshiba Corporation,Yokohama, Japan. 2) Power and Industrial Systems Research and Development Center, ToshibaCorporation, Yokohama, Japan. 3) Nuclear Technology Application Dept., Toshiba Nuclear EngineeringServices Corporation, Yokohama, JapanThe 4S is a sodium-cooled small fast reactor which is reflector-controlled for operationthrough core lifetime about 30 years. The nuclear design method has been selectedto treat neutron leakage with high accuracy. It consists of a continuous-energy MonteCarlo code, discrete ordinate transport codes and JENDL-3.3. These two types ofneutronic analysis codes are used for the design in a complementary manner. The accuracyof the codes has been evaluated by analysis of benchmark critical experimentsand the experimental reactor data. The measured data used for the evaluation is criticalexperimental data of the FCA XXIII as a physics mockup assembly of the 4S core,FCA XVI, FCA XIX, ZPR, and data of experimental reactor JOYO MK-1. Evaluatedcharacteristics are criticality, reflector reactivity worth, power distribution, absorber reactivityworth, and sodium void worth. A multi-component bias method was applied,especially to improve the accuracy of sodium void reactivity worth. As the result, it hasbeen confirmed that the 4S core nuclear design method provides good accuracy, andtypical bias factors and their uncertainties are determined.2:20 PMComparison of a Nuscale SMR Conceptual Core Design UsingCASMO5/SIMULATE5 and MCNP5Brandon Haugh (1), Ali Mohamed (2)1) Studsvik Scandpower Inc., Wilmington, NC. 2) NuScale Power Inc., Corvallis, ORA key issue during the initial startups of new Small Modular Reactors (SMRs) is thelack of operational data for reactor model validation. To help better understand theaccuracy of the reactor analysis codes CASMO5 and SIMULATE5, higher order comparisonsto MCNP5 have been performed. These comparisons are for an initial coreconceptual design of the NuScale reactor. The data have been evaluated at Hot ZeroPower (HZP) conditions. Comparisons of core reactivity, fuel temperature coefficient(FTC), and moderator temperature coefficients (MTC) have been performed. Comparisonresults show good agreement between CASMO5/SIMULATE5 and MCNP5for the conceptual initial core design.2:45 PMAdvanced High-Temperature Reactor Neutronic Core DesignDan Ilas, David E. Holcomb, and Venugopal K. VarmaOak Ridge National Laboratory, Oak Ridge, TN, USAThe Advanced High-Temperature Reactor (AHTR) is a 3400 MWth fluoride-salt-cooledhigh-temperature class design concept intended to serve as a central generating stationtype power plant. This paper focuses on preliminary neutronic design studies performedat Oak Ridge National Laboratory (ORNL) during fiscal year 2011. After a briefpresentation of the AHTR design concept, the paper summarizes several neutronicstudies. An optimization study for the AHTR core is first presented, and a design thatreaches a single-batch fuel length of just over 2 years is chosen as reference. Thetemperature and void coefficients of reactivity are then analyzed for a few configurationsof interest, and, in all cases analyzed, the isothermal temperature coefficientis proved to be negative. A discussion of the limiting factors due to the fast neutronfluence follows. The neutronic studies conclude with a discussion of the control andshutdown options. The studies presented confirm that sound neutronic alternatives existfor the design of the AHTR to maintain full passive safety features and reasonableoperation conditions. While significant technology development and demonstrationremain, the basic design concept appears sound and tolerant of much of the remainingperformance uncertainty. No fundamental impediments have been identified thatwould prevent widespread deployment of the concept.3:10 PMRadiogenic Lead with Dominant Content Of 208Pb: New Coolantand Neutron Moderator for Innovative Nuclear ReactorsShmelev A.N., Kulikov G.G., Kryuchkov E.F., Apse V.A., Kulikov E.G.National Research Nuclear University “MEPhI”, Moscow, RussiaThe advantages of radiogenic lead with dominant content of 208Pb as a reactor coolantwith respect to natural lead are caused by unique nuclear properties of 208Pbwhich is a double-magic nucleus with closed proton and neutron shells. This resultsin significantly lower micro cross section and resonance integral of radiative neutroncapture by 208Pb than those for numerous light neutron moderators. The extremelyweak ability of 208Pb to absorb neutrons results in the following effects. Firstly, neutronmoderating factor (ratio of scattering to capture cross sections) is larger than thatfor graphite and light water. Secondly, age and diffusion length of thermal neutrons arelarger than those for graphite, light and heavy water. Thirdly, neutron lifetime in 208Pbis comparable with that for graphite, beryllium and heavy water what could be importantfor safe reactor operation. The paper presents some results obtained in neutronicsand thermal-hydraulics evaluations of the benefits from the use of radiogenic leadwith dominant content of 208Pb instead of natural lead as a coolant of fast breederreactors. The paper demonstrates that substitution of radiogenic lead for natural leadcan offer the following benefits for operation of fast breeder reactors. Firstly, improvementof the reactor safety thanks to the better values of coolant temperature reactivitycoefficient and, secondly, improvement of some thermal-hydraulic reactor parameters.Radiogenic lead can be extracted from thorium sludge without isotope separation as208Pb is a final isotope in the decay chain of 232Th.17

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationMonday April 16, 2012 - 1:30 PM - 301 DSS2A - Radiation Transport Methods for Whole Reactor Core AnalysisSession Chair: A. Haghighat (VT); Tom Sutton (KAPL)1:30 PMOverlapping Local/Global Iteration Framework for Whole-Core Transport SolutionNam Zin Cho, Seungsu Yuk, Han Jong Yoo, and Sunghwan YunKorea Advanced Institute of Science and Technology (KAIST), Daejeon, KoreaIn current practice of reactor design analysis, whole-core diffusion nodal method isused in which nodal parameters are provided by single-assembly lattice physics calculationwith net current zero boundary condition. Thus, the whole-core solution isnot transport, because the inter-assembly transport effect is not incorporated. In thispaper, the overlapping local/global iteration framework is described that removes thelimitation of the current method. It consists of two-level iterative computations: halfnodeoverlapping local problems embedded in a global problem. The local problemcan employ fine-group deterministic or continuous-energy stochastic (Monte Carlo)transport methods, while the global problem is an equivalent coarse-group transportmodel based on p-CMFD methodology. The method is tested on several highly heterogeneousmulti-slab problems with encouraging results.1:55 PMThe Comet Method in 3-D Hexagonal GeometryKevin John Connolly and Farzad RahnemaNuclear and Radiological Engineering and Medical Physics Programs, George W. Woodruff School,Georgia Institute of Technology, Atlanta, Georgia, USAThe hybrid stochastic-deterministic coarse mesh radiation transport (COMET) methoddeveloped at Georgia Tech now solves reactor core problems in 3-D hexagonal geometry.In this paper, the method is used to solve three preliminary test problemsdesigned to challenge the method with steep flux gradients, high leakage, and strongasymmetry and heterogeneity in the core. The test problems are composed of blockstaken from a high temperature test reactor benchmark problem. As the method is stillin development, these problems and their results are strictly preliminary. Results arecompared to whole core Monte Carlo reference solutions in order to verify the method.Relative errors are on the order of 50 pcm in core eigenvalue, and mean relative errorin pin fission density calculations is less than 1% in these difficult test cores. Themethod requires the one-time pre-computation of a response expansion coefficientlibrary, which may be compiled in a comparable amount of time to a single whole coreMonte Carlo calculation. After the library has been computed, COMET may solve anynumber of core configurations on the order of an hour, representing a significant gainin efficiency over other methods for whole core transport calculations.2:20 PMFull Core Reactor Analysis: Running Denovo on JaguarJoshua J. Jarrell, Andrew T. Godfrey, Thomas M. Evans, and Gregory G.DavidsonOak Ridge National Laboratory, Oak Ridge, TNFully-consistent, full-core, 3D, deterministic neutron transport simulations using theorthogonal mesh code Denovo were run on the massively parallel computing architectureJaguar XT5. Using energy and spatial parallelization schemes, Denovo was ableto efficiently scale to more than 160k processors. Cell-homogenized cross sectionswere used with step-characteristics, linear-discontinuous finite element, and trilineardiscontinuousfinite element spatial methods. It was determined that using the finiteelement methods gave considerably more accurate eigenvalue solutions for largeaspectratio meshes than using step-characteristics.2:45 PMCoupled Monte Carlo Neutronics and Thermal Hydraulics forPower ReactorsW. Bernnat, M. Buck, M. Mattes (1), W. Zwermann, I. Pasichnyk, K. Velkov(2)1) Institut für Kernenergetik und Energiesysteme (IKE), Universität Stuttgart, Stuttgart, Germany. 2) Gesellschaftfür Anlagen- und Reaktorsicherheit (GRS) mbH, Garching, GermanyThe availability of high performance computing resources enables more and more theuse of detailed Monte Carlo models even for full core power reactors. The detailedstructure of the core can be described by lattices, modeled by so-called repeatedstructures e.g. in Monte Carlo codes such as MCNP5 or MCNPX. For cores with mainlyuniform material compositions, fuel and moderator temperatures, there is no problemin constructing core models. However, when the material composition and the temperaturesvary strongly a huge number of different material cells must be described whichcomplicate the input and in many cases exceed code or memory limits. The secondproblem arises with the preparation of corresponding temperature dependent crosssections and thermal scattering laws. Only if these problems can be solved, a realisticcoupling of Monte Carlo neutronics with an appropriate thermal-hydraulics model ispossible. In this paper a method for the treatment of detailed material and temperaturedistributions in MCNP5 is described based on user-specified internal functionswhich assign distinct elements of the core cells to material specifications (e.g. waterdensity) and temperatures from a thermal-hydraulics code. The core grid itself can bedescribed with a uniform material specification. The temperature dependency of crosssections and thermal neutron scattering laws is taken into account by interpolation,requiring only a limited number of data sets generated for different temperatures. Applicationswill be shown for the stationary part of the Purdue PWR benchmark usingATHLET for thermal- hydraulics and for a generic Modular High Temperature reactorusing THERMIX for thermal- hydraulics.3:10 PMComet Whole-Core Solution to a Stylized 3-Dimensional PressurizedWater Reactor Benchmark Problem With UO2 AndMOX FuelDingkang Zhang and Farzad RahnemaGeorgia Institute of Technology, Atlanta, GAA stylized pressurized water reactor (PWR) benchmark problem with UO2 and MOXfuel was used to test the accuracy and efficiency of the coarse mesh radiation transport(COMET) code. The benchmark problem contains 125 fuel assemblies and 44,000fuel pins. The COMET code was used to compute the core eigenvalue and assemblyand pin power distributions for three core configurations. In these calculations, a setof tensor products of orthogonal polynomials were used to expand the neutron angularphase space distribution on the interfaces between coarse meshes. The COMETcalculations were compared with the Monte Carlo code MCNP reference solutionsusing a recently published an 8-group material cross section library. The comparisonshowed both the core eigenvalues and assembly and pin power distributions predicatedby COMET agree very well with the MCNP reference solution if the orders of theangular flux expansion in the two spatial variables and the polar and azimuth angleson the mesh boundaries are 4, 4, 2 and 2. The mean and maximum differences in thepin fission density distribution ranged from 0.28%-0.44% and 3.0%-5.5%, all within3-sigma uncertainty of the MCNP solution. These comparisons indicate that COMETcan achieve accuracy comparable to Monte Carlo. It was also found that COMET’scomputational speed is 450 times faster than MCNP.18

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationMonday April 16, 2012 - 1:30 PM - 301 E8A - Advanced Modeling & Simulation in Reactor PhysicsSession Chair: Tom Downar (U. Mich.); Bill Martin (U. Mich.)1:30 PMEnhancements in SCALE 6.1B. T. Rearden, L. M. Petrie, D. E. Peplow, M. A. Jessee, D. Wiarda, M. L.Williams, R. A. Lefebvre, J. P. Lefebvre, I. C. Gauld, and S. GoluogluOak Ridge National Laboratory, Oak Ridge, TennesseeThe SCALE code system developed at Oak Ridge National Laboratory provides acomprehensive, verified and validated, user-friendly tool set for criticality safety, reactorphysics, radiation shielding, and sensitivity and uncertainty analysis. For more than30 years, regulators, licensees, and research institutions around the world have usedSCALE for nuclear safety analysis and design. SCALE provides a “plug-and-play”framework with 89 computational modules, including three deterministic and threeMonte Carlo radiation transport solvers that are selected based on the desired solution.SCALE’s graphical user interfaces assist with accurate system modeling, visualization,and convenient access to desired results. SCALE 6.1 builds on the existingcapabilities and ease-of-use of SCALE and provides several new features such asenhanced lattice physics capabilities and multigroup Monte Carlo depletion, improvedoptions and capabilities for sensitivity and uncertainty analysis calculations, improvedflexibility in shielding and criticality accident alarm system calculations with automatedvariance reduction, and new options for the definition of group structures for depletioncalculations. The SCALE 6.1 development team has focused on improved robustnessvia substantial additional regression testing and verification for new and existingfeatures, providing improved performance relative to SCALE 6.0, especially in reactorphysics calculations and in the nuclear data used for source term characterization andshielding calculations.1:55 PMALEPH22 – A General Purpose Monte Carlo Depletion CodeA.Stankovskiy and G.Van den Eynde, and P. Baeten (1), C.Trakas, P.-M.Demy and L.Villatte (2)1) SCK•CEN, Mol, Belgium, 2) AREVA NP, Tour AREVA, Pl. Paris La Défense, FranceThe Monte-Carlo burn-up code ALEPH is being developed at SCK•CEN since 2004.A previous version of the code implemented the coupling between the Monte Carlotransport (any version of MCNP or MCNPX) and the “deterministic” depletion codeORIGEN-2.2 but had important deficiencies in nuclear data treatment and limitationsinherent to ORIGEN-2.2. A new version of the code, ALEPH2, has several unique featuresmaking it outstanding among other depletion codes. The most important featureis full data consistency between steady-state Monte Carlo and time-dependent depletioncalculations. The last generation general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII and JENDL-4) are fully implemented, including special purposeactivation, spontaneous fission, fission product yield and radioactive decay data. Thebuilt-in depletion algorithm allows to eliminate the uncertainties associated with obtainingthe time-dependent nuclide concentrations. A predictor-corrector mechanism,calculation of nuclear heating, calculation of decay heat, decay neutron sources areavailable as well. The validation of the code on the results of REBUS experimentalprogram has been performed. The ALEPH2 has shown better agreement with measureddata than other depletion codes.2:20 PMMonte Carlo Depletion Calculations Using VESTA 2.1 NewFeatures and PerspectivesW. Haeck, B. Cochet and L. AguiarInstitut de Radioprotection et de Sûreté Nucléaire (IRSN), Fontenay-aux-Roses Cedex, FranceVESTA is a Monte Carlo depletion interface code that is currently under developmentat IRSN. With VESTA, the emphasis lies on both accuracy and performance, so thatthe code will be capable of providing accurate and complete answers in an acceptableamount of time compared to other Monte Carlo depletion codes. From its inception,VESTA is intended to be a generic interface code so that it will ultimately be capableof using any Monte-Carlo code or depletion module and that can be tailored to theusers needs. A new version of the code (version 2.1.x) will be released in 2012. Themost important additions to the code are a burn up dependent isomeric branchingratio treatment to improve the prediction of metastable nuclides such as 242mAm andthe integration of the PHOENIX point depletion module (also developed at IRSN) toovercome some of the limitations of the ORIGEN 2.2 module. The task of extractingand visualising the basic results and also the calculation of physical quantities or otherdata that can be derived from the basic output provided by VESTA will be the taskof the AURORA depletion analysis tool which will be released at the same time asVESTA 2.1.x. The experimental validation database was also extended for this newversion and it now contains a total of 35 samples with chemical assay data and 34assembly decay heat measurements.2:45 PMCoupled Full Core Neutron Transport/CFD Simulations ofPressurized Water ReactorsBrendan Kochunas, Shane Stimpson, Benjamin Collins, and Thomas Downar(1), Robert Brewster and Emilio Baglietto (2), Jin Yan (3)1) Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI.2) CD-adapco, Melville, NY. 3) Westinghouse Electric Company LLC, Columbia, SC, USARecently as part of the CASL project, a capability to perform 3D whole-core coupledneutron transport and computational fluid dynamics (CFD) calculations was demonstrated.This work uses the 2D/1D transport code DeCART and the commercial CFDcode STAR-CCM+. It builds on previous CASL work demonstrating coupling for smallerspatial domains. The coupling methodology is described along with the problemsimulated and results are presented for fresh hot full power conditions. An additionalcomparison is made to an equivalent model that uses lower order T/H feedback toassess the importance and cost of high fidelity feedback to the neutronics problem. Asimulation of a quarter core Combustion Engineering (CE) PWR core was performedwith the coupled codes using a Fixed Point Gauss-Seidel iteration technique. Thetotal approximate calculation requirements are nearly 10,000 CPU hours and 1 TB ofmemory. The problem took 6 coupled iterations to converge. The CFD coupled modeland low order T/H feedback model compared well for global solution parameters, witha difference in the critical boron concentration and average outlet temperature of 14ppmB and 0.94 °C, respectively. Differences in the power distribution were more significantwith maximum relative differences in the core-wide pin peaking factor (Fq) of5.37% and average relative differences in flat flux region power of 11.54%. Futurework will focus on analyzing problems more relevant to CASL using models with lessapproximations.3:10 PMPacking Microstructure and Local Density Variations of Experimentaland Computational Pebble BedsG. J. Auwerda, J. L. Kloosterman, D. Lathouwers and T. H. J. J. van derHagenDelft University of Technology, Delft, The NetherlandsIn pebble bed type nuclear reactors the fuel is contained in graphite pebbles, whichform a randomly stacked bed with a non-uniform packing density. These variations caninfluence local coolant flow and power density and are a possible cause of hotspots. Toanalyse local density variations computational methods are needed that can generaterandomly stacked pebble beds with a realistic packing structure on a pebble-to-pebblelevel. We first compare various properties of the local packing structure of a computedbed with those of an image made using computer aided X-ray tomography, looking atproperties in the bulk of the bed and near the wall separately. Especially for the bulkof the bed, properties of the computed bed show good comparison with the scannedbed and with literature, giving confidence our method generates beds with realisticpacking microstructure. Results also show the packing structure is dierent near thewall than in the bulk of the bed, with pebbles near the wall forming ordered layerssimilar to hexagonal close packing. Next, variations in the local packing density areinvestigated by comparing probability density functions of the packing fraction of smallclusters of pebbles throughout the bed. Especially near the wall large variations in localpacking fractions exists, with a higher probability for both clusters of pebbles withlow (0.65) packing fraction, which could significantly aect flow ratesand, together with higher power densities, could result in hotspots.19

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationMonday April 16, 2012 - 3:50 PM - 301 A1B - Core Analysis MethodsSession Chair: Brian Elder (TVA); W. Zwermann (GRS)3:50 PMDevelopment of 3D Pseudo Pin-by-Pin Calculation Methodologyin ANCBaocheng Zhang, Larry Mayhue, Harish Huria, and Boyan Ivanov.Westinghouse Electric Company LLC, Cranberry, PA, USAAdvanced cores and fuel assembly designs have been developed to improve operationalflexibility, economic performance and further enhance safety features of nuclearpower plants. The simulation of these new designs, along with strong heterogeneousfuel loading, have brought new challenges to the reactor physics methodologies currentlyemployed in the industrial codes for core analyses. Control rod insertion duringnormal operation is one operational feature in the AP1000® plant of Westinghousenext generation Pressurized Water Reactor (PWR) design. This design improves itsoperational flexibility and efficiency but significantly challenges the conventional reactorphysics methods, especially in pin power calculations. The mixture loading of fuelassemblies with significant neutron spectrums causes a strong interaction betweendifferent fuel assembly types that is not fully captured with the current core designcodes. To overcome the weaknesses of the conventional methods, Westinghouse hasdeveloped a state-of-the-art 3D Pin-by-Pin Calculation Methodology (P3C) and successfullyimplemented in the Westinghouse core design code ANC. The new methodologyhas been qualified and licensed for pin power prediction. The 3D P3C methodologyalong with its application and validation will be discussed in the paper.4:15 PMWIMS/PANTHER Analysis of UO2/MOX Cores Using EmbeddedSupercellsMartin Knight and Paul Bryce (1), Sheldon Hall (2)1) EDF Energy, Barnwood, Gloucester, UK. 2) Advanced Modelling and Computation Group, ImperialCollege, London, UKThis paper describes a method of analysing PWR UO2/MOX cores with WIMS/PAN-THER. Embedded supercells, run within the reactor code, are used to correct thestandard methodology of using 2-group smeared data from single assembly latticecalculations. In many other codes the weakness of this standard approach has beenimproved for MOX by imposing a more realistic environment in the lattice code, or byimproving the sophistication of the reactor code. In this approach an intermediate setof calculations is introduced, leaving both lattice and reactor calculations broadly unchanged.The essence of the approach is that the whole core is broken down into a setof “embedded” supercells, each extending over just four quarter assemblies, with zeroleakage imposed at the assembly mid-lines. Each supercell is solved twice, first witha detailed multi-group pin-by-pin solution, and then with the standard single assemblyapproach. Correction factors are defined by comparing the two solutions, and thesecan be applied in whole core calculations. The restriction that all such calculations aremodelled with zero leakage means that they are independent of each other and of thecore-wide flux shape. This allows parallel pre-calculation for the entire cycle once theloading pattern has been determined, in much the same way that single assemblylattice calculations can be pre-calculated once the range of fuel types is known. Comparisonsagainst a whole core pin-by-pin reference demonstrates that the embeddingprocess does not introduce a significant error, even after burnup and refuelling. Comparisonsagainst a WIMS reference demonstrate that a pin-by-pin multi-group diffusionsolution is capable of capturing the main interface effects. This therefore definesa practical approach for achieving results close to lattice code accuracy, but broadly atthe cost of a standard reactor calculation.4:40 PMEvaluation Of Rodded BWR Assembly Pin Powers With SIMU-LATE5Tamer Bahadir (1), Sten-Örjan Lindahl (2)1) Studsvik Scandpower, Inc. Waltham, MA. 2) Studsvik Scandpower AB, Västerås, SwedenIn the development of Studsvik’s nodal code SIMULATE5, special attention has beengiven to the accurate description of pin powers. The code solves the multi-group diffusionor simplified P3 equations with high spatial resolution. A short description isgiven of the handling of material heterogeneities in the axial and radial directions – theaxial re-homogenization and the radial submesh model, respectively. Furthermore,two models which are of special importance in the presence of control rods, and henceon the pellet clad interaction (PCI) phenomenon, are described; the quarter-assemblythermal-hydraulic treatment and the depletion of absorber material. A numerical examplefor a real core shows that while the impact of the fine grained description isnegligible on global parameters such as keff, the effect on pin powers may be substantial.For a deeply inserted and highly depleted control rod, the difference in pin powerscaused by neglecting the advanced models of SIMULATE5, may be as high as 15 %.5:05 PMExperimental Validation of 3D Reconstructed Pin-Power Distributionsin Full-Scale BWR Fuel Assemblies with PartialLength RodsFlavio Dante Giust (1), Peter Grimm and Rakesh Chawla (2)1) Axpo Kernenergie, CH-5401 Baden, Switzerland. 2) Paul Scherrer Institute, Villigen PSI, SwitzerlandTotal fission rate measurements have been performed on full-size BWR fuel assembliesof type SVEA-96 Optima2 in the framework of Phase III of the LWR-PROTEUSexperimental program at the Paul Scherrer Institute. This paper presents comparisonsof calculated, nodal reconstructed, pin-wise total-fission rate distributions with experimentalresults. Radial comparisons have been performed for the three sections of theassembly (96, 92 and 84 fuel pins), while three-dimensional effects have been investigatedat pellet-level for the two transition regions, i.e. the tips of the short (1/3) andlong (2/3) partial length rods. The test zone has been modeled using two different codesystems: HELIOS/PRESTO-2 and CASMO-5/SIMULATE-5. The first is presently usedfor core monitoring and design at the Leibstadt Nuclear Power Plant (KKL). The secondrepresents the most recent generation of the widely applied CASMO/SIMULATEsystem. For representing the PROTEUS test-zone boundaries, Partial Current Ratios(PCRs) – derived from a 3D MCNPX model of the entire reactor – have been applied tothe PRESTO-2 and SIMULATE-5 models in the form of 2- and 5-group diagonal albedomatrices, respectively. The MCNPX results have also served as a reference, highordertransport solution in the calculation/experiment comparisons. It is shown thatthe performance of the nodal methodologies in predicting the global distribution of thetotal-fission rate is very satisfactory. Considering the various radial comparisons, thestandard deviations of the calculated/experimental (C/E) distributions do not exceed1.9% for any of the three methodologies – PRESTO-2, SIMULATE-5 and MCNPX. Forthe three-dimensional comparisons at pellet-level, the corresponding standard deviationsare 2.7%, 2.0% and 2.1%, respectively.5:30 PMPost-refinement Multiscale Method for Pin Power ReconstructionBenjamin Collins, Volkan Seker, and Thomas DownarDepartment of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MIThe ability to accurately predict local pin powers in nuclear reactors is necessary to understandthe mechanisms that cause fuel pin failure during steady state and transientoperation. In the research presented here, methods are developed to improve the localsolution using high order methods with boundary conditions from a low order global solution.Several different core configurations were tested to determine the improvementin the local pin powers compared to the standard techniques based on diffusion theoryand pin power reconstruction (PPR). The post-refinement multiscale methods use theglobal solution to determine boundary conditions for the local solution. The local solutionis solved using either a fixed boundary source or an albedo boundary condition;this solution is “post-refinement” and thus has no impact on the global solution.20

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationMonday April 16, 2012 - 3:50 PM - 301 B2B - Deterministic Transport TheorySession Chair: Lei Zhu (Studsvik Scandpower); C. Rabiti (INL)3:50 PMSurface Harmonics Method Equations for Solving The Time-Dependent Neutron Transport Problems and Their VerificationV.F. Boyarinov, A.E. Kondrushin, P.A. FomichenkoNational Research Center, “Kurchatov Institute”, Moscow, RussiaFinite-difference time-dependent equations of Surface Harmonics method have beenobtained for plane geometry. Verification of these equations has been carried out bycalculations of tasks from “Benchmark Problem Book ANL-7416”. The capacity and efficiencyof the Surface Harmonics method have been demonstrated by solution of thetime-dependent neutron transport equation in diffusion approximation. The results ofstudies showed that implementation of Surface Harmonics method for full-scale calculationswill lead to a significant progress in the efficient solution of the time-dependentneutron transport problems in nuclear reactors.4:15 PMMixed Legendre Moments and Discrete Scattering Cross Sectionsfor Anisotropy RepresentationA. Calloo, J. F. Vidal, R. Le Tellier, G. RimpaultCEA, DEN, DER/SPRC/LEPh, Saint-Paul-lez-Durance, FranceThis paper deals with the resolution of the integro-differential form of the Boltzmanntransport equation for neutron transport in nuclear reactors. In multigroup theory,deterministic codes use transfer cross sections which are expanded on Legendrepolynomials. This modelling leads to negative values of the transfer cross section forcertain scattering angles, and hence, the multigroup scattering source term is wronglycomputed. The first part compares the convergence of “Legendre-expanded” crosssections with respect to the order used with the method of characteristics (MOC)for Pressurised Water Reactor (PWR) type cells. Furthermore, the cross section isdeveloped using piecewise-constant functions, which better models the multigrouptransfer cross section and prevents the occurrence of any negative value for it. Thesecond part focuses on the method of solving the transport equation with the abovementionedpiecewise-constant cross sections for lattice calculations for PWR cells.This expansion thereby constitutes a “reference” method to compare the conventionalLegendre expansion to, and to determine its pertinence when applied to reactor physicscalculations.4:40 PMDistributed Resonance Self-Shielding Using the EquivalencePrincipleDimitar AltiparmakovAtomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, CanadaThis paper presents an extension of the equivalence principle to allow distributed resonanceselfshielding in a multi-region fuel configuration. Rational expansion of fuel-tofuelcollision probability is applied in order to establish equivalence between the actualfuel configuration and a homogeneous mixture of hydrogen and resonant absorber,which is a commonly used model to calculate library tables of resonance integrals.The main steps in derivation are given along with the basic physics assumptions onwhich the presented approach relies. The method has been implemented in the latticecode WIMS-AECL and routinely used for calculation of CANDU-type reactor lattices.Its capabilities are illustrated by comparison of WIMS-AECL and MCNP results of238U resonance capture in a CANDU lattice cell. In order to determine optimal rationalexpansion of fuel-to-fuel collision probability, the calculations were carried out byvarying the number of rational terms from 1 to 6. The results show that 4 terms are sufficient.The further increase of the number of terms affects the computing time, whilethe impact on accuracy is negligible. To illustrate the convergence of the results, thefuel subdivision is gradually refined varying the number of fuel pin subdivisions from1 to 32 equal-area annuli. The results show very good agreement with the referenceMCNP calculation.5:05 PMDerivation of New 3D Discrete Ordinate EquationsCory D. AhrensColorado School of Mines, Department of Applied Mathematics and Statistics, Golden, ColoradoThe Sn equations have been the workhorse of deterministic radiation transport calculationsfor many years. Here we derive two new angular discretizations of the 3Dtransport equation. The first set of equations, derived using Lagrange interpolation andcollocation, retains the classical Sn structure, with the main difference being how thescattering source is calculated. Because of the formal similarity with the classical Snequations, it should be possible to modify existing computer codes to take advantageof the new formulation. In addition, the new Sn-like equations correctly capture deltafunction scattering. The second set of equations, derived using a Galerkin technique,does not retain the classical Sn structure because the streaming term is not diagonal.However, these equations can be cast into a form similar to existing methods developedto reduce ray effects. Numerical investigation of both sets of equations is underway.5:30 PMIsogeometric Analysis for Neutron Diffusion ProblemsS.K. Hall, M.D. Eaton and M.M.R.WilliamsDepartment of Earth Science and Engineering, Imperial College, South Kensington Campus, London,UKIsogeometric analysis can be viewed as a generalisation of the finite element method.It permits the exact representation of a wider range of geometries including conic sections.This is possible due to the use of concepts employed in computer-aided design.The underlying mathematical representations from computer-aided design are usedto capture both the geometry and approximate the solution. In this paper the neutrondiffusion equation is solved using isogeometric analysis. The practical advantages arehighlighted by looking at the problem of a circular fuel pin in a square moderator. Forthis problem the finite element method requires the geometry to be approximated. Thisleads to errors in the shape and size of the interface between the fuel and the moderator.In contrast to this isogeometric analysis allows the interface to be representedexactly. It is found that, due to a cancellation of errors, the finite element methodconverges more quickly than isogeometric analysis for this problem. A fuel pin in avacuum was then considered as this problem is highly sensitive to the leakage acrossthe interface. In this case isogeometric analysis greatly outperforms the finite elementmethod. Due to the improvement in the representation of the geometry isogeometricanalysis can outperform traditional finite element methods. It is proposed that the useof isogeometric analysis on neutron transport problems will allow deterministic solutionsto be obtained for exact geometries. Something that is only currently possiblewith Monte Carlo techniques.21

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationMonday April 16, 2012 - 3:50 PM - 301 C3A - Monte Carlo Methods & DevelopmentsSession Chair: Bill Martin (U. Mich.)3:50 PMComparison of Discrete and Continuous Thermal NeutronScattering Treatments In MCNP5A.T. Pavlou (1), F.B. Brown (2), W.R. Martin (1), B.C. Kiedrowski (2)1) University of Michigan, Department of Nuclear Engineering & Radiological Sciences, Ann Arbor, MI,USA. 2) Los Alamos National Laboratory, Monte Carlo Codes Group, Los Alamos, NM, USAThe standard discrete thermal neutron S(α,β) scattering treatment in MCNP5 is comparedwith a continuous S(α,β) scattering treatment using a criticality suite of 119benchmark cases and ENDF/B-VII.0 nuclear data. In the analysis, six bound isotopesare considered: beryllium metal, graphite, hydrogen in water, hydrogen in polyethylene,beryllium in beryllium oxide and oxygen in beryllium oxide. Overall, there are onlysmall changes in the eigenvalue (keff) between discrete and continuous treatments.In the comparison of 64 cases that utilize S(α,β) scattering, 62 agreed at the 95% confidencelevel, and the 2 cases with differences larger than 3 σ agreed within 1 σ whenmore neutrons were run in the calculations. The results indicate that the changes ineigenvalue between continuous and discrete treatments are random, small, and wellwithin the uncertainty of measured data for reactor criticality experiments.4:15 PMTemperature Effects of Resonance Scattering for EpithermalNeutrons in MCNPE.E. Sunny (1), F.B. Brown, B.C. Kiedrowski (2), W.R. Martin (1)1) University of Michigan, Department of Nuclear Engineering & Radiological Sciences, Ann Arbor, MI,USA. 2) Los Alamos National Laboratory, Monte Carlo Codes Group, Los Alamos, NM, USAEpithermal neutron elastic scattering can be significantly affected by the thermal motionof target nuclides. Since the 1950s continuous-energy Monte Carlo codes havegenerally accounted for the target motion using a free gas scattering model, with theassumption that the scattering crosssection is constant in energy. Recent work hasshown the importance of resonance scattering, and several methods for an improvedfree-gas treatment have been developed. We have implemented a rejection-basedsampling scheme in the MCNP free-gas treatment to account for cross-section variation.The modified MCNP code was used to investigate a number of practical concerns:results for an LWR Doppler defect benchmark; computational costs; and energylimits for the freegas treatment. Additionally, the impact on a suite of ICSBEP criticalitybenchmark problems (at room temperature) was determined to be negligible, animportant result since such problems are used extensively in testing and evaluatingrevisions to ENDF/B-VII nuclear data.4:40 PMGeneral Purpose Dynamic Monte Carlo with Continuous Energyfor Transient AnalysisBart L. Sjenitzer and J. Eduard HoogenboomDelft University of Technology, Department of Radiation, Radionuclide and Reactors, Delft, the NetherlandsFor safety assessments transient analysis is an important tool. It can predict maximumtemperatures during regular reactor operation or during an accident scenario.Despite the fact that this kind of analysis is very important, the state of the art still usesrather crude methods, like diffusion theory and point-kinetics. For reference calculationsit is preferable to use the Monte Carlo method. In this paper the dynamic MonteCarlo method is implemented in the general purpose Monte Carlo code Tripoli4. Also,the method is extended for use with continuous energy. The first results of DynamicTripoli demonstrate that this kind of calculation is indeed accurate and the results areachieved in a reasonable amount of time. With the method implemented in Tripoli it isnow possible to do an exact transient calculation in arbitrary geometry.5:05 PMRobust Volume Calculations for Constructive Solid Geometry(CSG) Components in Monte Carlo Transport CalculationsDavid L. Millman, David P. Griesheimer, Brian R. Nease, and Jack SnoeyinkDepartment of Computer Science, University of North Carolina at Chapel Hill and Bechtel Marine PropulsionCorporation, Bettis Atomic Power LaboratoryIn this paper we consider a new generalized algorithm for the efficient calculation ofcomponent object volumes given their equivalent constructive solid geometry (CSG)definition. The new method relies on domain decomposition to recursively subdividethe original component into smaller pieces with volumes that can be computed analyticallyor stochastically, if needed. Unlike simpler brute-force approaches, the proposeddecomposition scheme is guaranteed to be robust and accurate to within auser-defined tolerance. The new algorithm is also fully general and can handle anyvalid CSG component definition, without the need for additional input from the user.The new technique has been specifically optimized to calculate volumes of componentdefinitions commonly found in models used for Monte Carlo particle transport simulationsfor criticality safety and reactor analysis applications. However, the algorithm canbe easily extended to any application which uses CSG representations for componentobjects. The paper provides a complete description of the novel volume calculationalgorithm, along with a discussion of the conjectured error bounds on volumes calculatedwithin the method. In addition, numerical results comparing the new algorithmwith a standard stochastic volume calculation algorithm are presented for a series ofproblems spanning a range of representative component sizes and complexities.5:30 PMRevised Methods for Few-Group Cross Sections Generationin the Serpent Monte Carlo CodeEmil Fridman (1), Jaakko Leppänen (2)1) Reactor Safety Division Helmholz-Zentrum Dresden-Rossendorf, Dresden, Germany. 2) VTT TechnicalResearch Centre of Finland, FinlandThis paper presents new calculation methods, recently implemented in the SerpentMonte Carlo code, and related to the production of homogenized few-group constantsfor deterministic 3D core analysis. The new methods fall under three topics: 1) Improvedtreatment of neutron-multiplying scattering reactions, 2) Group constant generationin reflectors and other non-fissile regions and 3) Homogenization in leakage-correctedcriticality spectrum. The methodology is demonstrated by a numerical example,comparing a deterministic nodal diffusion calculation using Serpent-generated crosssections to a reference full-core Monte Carlo simulation. It is concluded that the newmethodology improves the results of the deterministic calculation, and paves the wayfor Monte Carlo based group constant generation.22

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationMonday April 16, 2012 - 3:50 PM - 301 DSS2B - Radiation Transport Methods for Whole Reactor Core AnalysisSession Chair: Tom Evans (ORNL); Josh Jarrell (ORNL)3:50 PMMC21 Analysis of the Nuclear Energy Agency Monte CarloPerformance Benchmark ProblemDaniel J. Kelly and Thomas M. Sutton (1), Stephen C. Wilson (2)1) Knolls Atomic Power Laboratory – Bechtel Marine Propulsion Corporation, Schenectady, NY. 2) BettisAtomic Power Laboratory – Bechtel Marine Propulsion Corporation, West Mifflin, PADue to the steadily decreasing cost and wider availability of large scale computingplatforms, there is growing interest in the prospects for the use of Monte Carlo forreactor design calculations that are currently performed using few-group diffusiontheory or other low-order methods. To facilitate the monitoring of the progress beingmade toward the goal of practical full-core reactor design calculations using MonteCarlo, a performance benchmark has been developed and made available throughthe Nuclear Energy Agency. A first analysis of this benchmark using the MC21 MonteCarlo code was reported on in 2010, and several practical difficulties were highlighted.In this paper, a newer version of MC21 that addresses some of these difficulties hasbeen applied to the benchmark. In particular, the confidence-interval-determinationmethod has been improved to eliminate source correlation bias, and a fission-sourceweightingmethod has been implemented to provide a more uniform distribution ofstatistical uncertainties. In addition, the Forward-Weighted, Consistent-Adjoint-DrivenImportance Sampling methodology has been applied to the benchmark problem. Resultsof several analyses using these methods are presented, as well as results froma very large calculation with statistical uncertainties that approach what is needed fordesign applications.4:15 PMWhole-Core COMET Solutions to A 3-Dimensional PWRBenchmark Problem With GadoliniumDingkang Zhang and Farzad RahnemaGeorgia Institute of Technology, Atlanta, GA, USAA pressurized water reactor (PWR) benchmark problem with Gadolinium was usedto determine the accuracy and computational efficiency of the coarse mesh radiationtransport method COMET. The benchmark problem contains 193 square fuel assemblies.The COMET solution (eigenvalue, assembly averaged and fuel pin averaged fissiondensity distributions) was compared with those obtained from the correspondingMonte Carlo reference solution using the same 2-group material cross section library.The comparison showed that both the core eigenvalue and fission density distributionaveraged over each assembly and fuel pin predicated by COMET agree very well withthe corresponding MCNP reference solution if the incident flux response expansionused in COMET is truncated at 2nd order in the two spatial and the two angular variables.The benchmark calculations indicate that COMET has Monte Carlo accuracy.In, particular, the eigenvalue difference between the codes ranged from 17 pcm to 35pcm, being within 2 standard deviations of the calculational uncertainty. The mean fluxweighted relative differences in the assembly and fuel pin fission densities were 0.47%and 0.65%, respectively. It was also found that COMET’s full (whole) core computationalspeed is 30,000 times faster than MCNP in which only 1/8th of the core is modeled.It is estimated that COMET would have been about over 6 orders of magnitudefaster than MCNP if the full core were also modeled in MCNP.4:40 PMAnalysis of the Pool Critical Assembly Pressure VesselBenchmark Using PentranChristopher A. Edgar and Glenn E. SjodenNuclear and Radiological Engineering Program, George W. Woodruff School of Mechanical Engineering,Georgia Institute of Technology, Atlanta, GAThe internationally circulated Pool Critical Assembly (PCA) Pressure Vessel Benchmarkwas analyzed using the PENTRAN Parallel Sn code system for the geometry,material, and source specifications as described in the PCA Benchmark documentation.This research focused on utilizing the BUGLE-96 cross section library and accompanyingreaction rates, while examining both adaptive differencing on a coarsemesh basis as well as Directional Theta Weighted Sn differencing in order to comparethe calculated PENTRAN results to measured data. The results show good comparisonwith the measured data as well as to the calculated results provided from TORTfor the BUGLE-96 cross sections and reaction rates, which suggests PENTRAN is aviable and reliable code system for calculation of light water reactor neutron shieldingand dosimetry calculations.5:05 PMCOMET Solutions to a Stylized BWR Benchmark ProblemDingkang Zhang and Farzad RahnemaGeorgia Institute of Technology, Atlanta, GA, USAIn this paper, a stylized 3-D BWR benchmark problem was used to evaluate the performanceof the coarse mesh radiation transport method COMET. The benchmark problemconsists of 560 fuel bundles at 3 different burnups and 3 coolant void states. TheCOMET solution was compared with the corresponding Monte Carlo reference solutionusing the same 2-group material cross section library for three control blade (rod)configurations, namely, all rods out (ARO), all rods in (ARI) and some rods in (SRI).The differences in the COMET and MCNP eigenvalues were 43 pcm, 66 pcm and 32pcm for the ARO, ARI and SRI cases, respectively. These differences are all within 3standard deviations of the COMET uncertainty. The average relative differences inthe bundle averaged fission densities for these three cases were 0.89%, 1.24%, and1.05%, respectively. The corresponding differences in the fuel pin averaged fissiondensities were 1.24%, 1.84% and 1.29%, respectively. It was found that COMET is3,000 times faster than Monte Carlo, while its statistical uncertainty in the fuel pin fissiondensity is much lower than that of Monte Carlo (i.e., ~40 times lower).5:30 PMAutomatic Whole Core Depletion & Criticality Calculations byMCNPX 2.7.0S. Kalcheva and E. KoonenSCK•CEN, BR2 Reactor Department, Mol, BelgiumDifferent approaches to perform automatic whole core criticality & depletion calculationsin a research reactor using MCNPX 2.7.0 are presented. An approximate methodis to use the existing symmetries of the burned fuel material distribution in thecore, i.e., the axial, radial and azimuth symmetries around the core center, in orderto significantly reduce the computation time. In this case it is not necessary to give aunique material number to each burn up cell. Cells having similar burn up and power,achieved during similar irradiation history at same initial fuel composition, will experiencesimilar composition evolution and can therefore be given the same material number.To study the impact of the number of unique burn up materials on the computationtime and utilized RAM memory, several MCNPX models have been developed. Thepaper discusses the accuracy of the model on comparison with measurements of BR2operation cycles in function of the number of unique burn up materials and the impactof the used Q-value (MeV/fission) of the recoverable fission energy.23

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationMonday April 16, 2012 - 3:50 PM - 301 E4B - Reactor Concepts & DesignsSession Chair: Kim Sissom (TVA); Sara Bortot (KTH)3:50 PMHigh Conversion Th-U233 fuel assembly for current generationof PWRsDaniela Baldova and Emil FridmanReactor Safety Division, Helmholtz-Zentrum Dresden-Rossendorf, Dresden, GermanyThis paper presents a preliminary design of a high conversion Th-U233 fuel assemblyapplicable for current generation of Pressurized Water Reactor (PWRs). The consideredfuel assembly has a typical 17×17 PWR lattice. However in order to increase theconversion of Th232 to U233, the assembly was subdivided into the two regions calledseed and blanket. The central seed region has a higher than blanket U233 content andacts as a neutron source for the peripheral blanket region. The latest acts as a U233breeder. While the seed fuel pins have a standard dimensions the blanket fuel radiuswas increased in order to reduce the moderation and to facilitate the resonance neutronabsorption in blanket Th232. The U233 content in the seed and blanket regionswas optimized to achieve maximal initial to discharged fissile inventory ratio (FIR)taking into account the target fuel cycle length of 12 months with 3-batch reloadingscheme. In this study the neutronic calculations were performed on the fuel assemblylevel using Helios deterministic lattice transport code. The fuel cycle length and thecore k-eff were estimated by applying the Non Linear Reactivity Model. The applicabilityof the HELIOS code for the analysis of the Th-based high conversion designswas confirmed with the help of continuous-energy Monte-Carlo code SERPENT. Theresults of optimization studies show that for the heterogeneous seed and blanket (SB)fuel assembly the FIR of about 0.95 can be achieved.4:15 PMNeutronic Analysis of a Fusion Hybrid ReactorT. KammashUniversity of Michigan, Ann Arbor, MIIn a physor 2010 paper(1), we introduced a fusion hybrid reactor whose fusion componentis the gasdynamic mirror (GDM), and whose blanket was made of thorium – 232.The thrust of that study was to demonstrate the performance of such a reactor by establishingthe breeding of uranium – 233 in the blanket, and the burning thereof to producepower. In that analysis, we utilized the diffusion equation for one-energy neutrongroup, namely, those produced by the fusion reactions, to establish the power distributionand density in the system. Those results should be viewed as a first approximationsince the high energy neutrons are not effective in inducing fission, but contributeprimarily to the production of actinides. In the presence of a coolant, however, such aswater, these neutrons tend to thermalize rather quickly, hence a better assessment ofthe reactor performance would require at least a two group analysis, namely the fastand thermal groups. We follow that approach and write an approximate set of equationsfor the fluxes of these groups. From these relations we deduce the all-importantquantity, keff, which we utilize to compute the multiplication factor, and subsequently,the power density in the reactor. We show that keff can be made to have a valueof 0.99, thus indicating that 100 thermal neutrons are generated per fusion neutron,while allowing the system to function as “subcritical.” Moreover, we show that such ahybrid reactor can generate hundreds of megawatts of thermal power per cm of lengthdepending on the flux of the fusion neutrons impinging on the blanket.4:40 PMAssessment of Possible Cycle Lengths for Fully Ceramic Micro-EncapsulatedFuel-Based Light Water Reactor ConceptsR. Sonat Sen, Michael A. Pope, Abderrafi M. Ougouag, Kemal Pasamehmetoglu(1), Francesco Venneri (2)1) Idaho National Laboratory, Idaho Falls, ID. 2) UltraSafe NuclearThe use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materialshas the potential to allow the design of extremely safe LWRs with accidenttolerantfuel. This paper examines the feasibility of LWR-like cycle length for such alow enriched uranium fuel with the imposed constraint of strictly retaining the originalgeometry of the fuel pins and assemblies. The motivation for retaining the originalgeometry is to provide the ability to incorporate the fuel “asis” into existing LWRswhile retaining their thermal-hydraulic characteristics. The feasibility of using this fuelis assessed by looking at cycle lengths and fuel failure rates. Other considerations(e.g., safety parameters, etc.) were not considered at this stage of the study. Thestudy includes the examination of different TRISO kernel diameters without changingthe coating layer thicknesses. The study shows that a naïve use of UO2 resultsin cycle lengths too short to be practical for existing LWR designs and operationaldemands. Increasing fissile inventory within the fuel compacts shows that acceptablecycle lengths can be achieved. In this study, starting with the recognized highest packingfraction practically achievable (44%), higher enrichment, larger fuel kernel sizes,and the use of higher density fuels have been evaluated. The models demonstratecycle lengths comparable to those of ordinary LWRs. As expected, TRISO particleswith extremely large kernels are shown to fail under all considered scenarios. In contrast,the designs that do not depart too drastically from those of the nominal NGNPHTR fuel TRISO particles are shown to perform satisfactorily and display a high rateof survival under all considered scenarios. Finally, it is recognized that relaxing thegeometry constraint will result in satisfactory cycle lengths even using UO2-loadedTRISO particles-based fuel with enrichment at or below 20 w/o.5:05 PMNuclear Design of Helical Cruciform Fuel RodsKoroush Shirvan and Mujid S. KazimiDepartment of Nuclear Engineering and Sciences, Massachusetts Institute of TechnologyIn order to increase the power density of current and new light water reactor designs,the Helical Cruciform Fuel (HCF) rods are proposed. The HCF rods are equivalentto a cylindrical rod, with the fuel in a cruciform shaped, twisted axially. The HCF rodsincrease the surface area to volume ratio and inter-subchannel mixing behavior dueto their cruciform and helical shapes, respectively. In a previous study, the HCF rodshave shown the potential to uprate existing PWRs by 50% and BWRs by 25%. However,HCF rods do display different neutronics modeling and performance. The cruciformcross section of HCF rods creates radially asymmetric heat generation andtemperature distribution. The nominal HCF rod’s beginning of life reactivity is reduced,compared to a cylindrical rod with the same fuel volume, by 500 pcm, due to increasein absorption in cladding. The rotation of these rods accounts for reactivity changes,which depends on the H/HM ratio of the pincell. The HCF geometry shows large sensitivitiesto U235 or gadolinium enrichments compared to a cylindrical geometry. Inaddition, the gadoliniumcontaining HCF rods show a stronger effect on neighboringHCF rods than in case of cylindrical rods, depending on the orientation of the HCFrods. The helical geometry of the rods introduces axial shadowing of about 600 pcm,not seen in typical cylindrical rods.5:30 PMPreliminary Design of Ultra-Long Cycle Fast Reactor EmployingBreed-and-Burn StrategyTae Woo Tak, Hwanyeal Yu, Ji Hyun Kim, and Deokjung Lee (1), T. K. Kim(2)1) Ulsan National Institute of Science and Technology, Ulsan, Korea. 2) Argonne National Laboratory,Argonne IL, USAA new design of ultra-long cycle fast reactor with power rate of 1000 MWe (UCFR)has been developed based on the strategy of breed-and burn. The bottom region ofthe core with low enriched uranium (LEU) plays a role of igniter of the core burningand the upper natural uranium (NU) region acts as blanket for breeding. Fissile materialsare bred in the blanket and the active core moves upward at a speed of 5.4 cm/year. Through the core depletion calculation using Monte Carlo code, McCARD, it isconfirmed that a full power operation of 60 years without refueling is feasible. Core performancecharacteristics have been evaluated in terms of axial/radial power shapes,reactivity feedback coefficients, etc. This design will serve as a base model for furtherdesign study of UCFRs using LWR spent fuels in the blanket region.24

Joan CronanWomen’s Athletic DirectorUniversity of TennesseePHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationMonday April 16, 2012 - 6:30 PM - Women’s Basketball Hall of FameBanquet at the Women’s Basketball Hall of FamePresentation by Joan Cronana at 7:30 PM“Building Successful Teams”Introduction by Peggy EmmettJoan Cronan is in her 29th year at the University of Tennessee where as the Women’s Athletic Director, shehas facilitated the operation of a first-class program, which has finished in the top two in the SoutheasternConference’s Women’s All-Sport Award from The New York Times for six of the last seven years. She has won numerous honors for herleadership including being named Athletic Director of the Year by the National Association of Collegiate Women Athletic Administrators in2005 and being honored with the Carl Maddox Sports Management Award by the U.S. Sports Academy in 2011. Under her leadership,the Lady Vols have been very successful in both the athletic and academic realms. Cronan is a graduate of LSU. She came to UT fromthe College of Charleston where she served as athletic director and also coached basketball and tennis. She is a past president of theTennessee Sports Hall of Fame and works closely with the Fellowship of Christian Athletes and Athletes in Action.25

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 8:00 AM - 301 A13A - Fuel, Material, Mechanical Analysis & BehaviorSession Chair: M. Defilippis (TVA); Lembit Sihver (Chalmers)8:00 AMFactors Influencing Helium Measurements for Detection 0fControl Rod Failures in BWRIrina Larsson and Lembit Sihver (1), Helena Loner and Guido Ledergerber(2), Bernhard Schnurr (3)1) Division of Nuclear Engineering, Department of Applied Physics, Chalmers University of Technology,Gothenburg, Sweden. 2) Kernkraftwerk Leibstadt, Leibstadt, Switzerland. 3) E.ON Kernkraft GmbH, Essenbach,GermanyMuch effort has been made to minimize the number and consequences of fuel failuresat nuclear power plants. The consequences of control rod failures have also gainedan increased attention. In this paper we introduce a system for on-line surveillanceof control rod integrity which has several advantages comparing to the surveillancemethods available today in boiling water reactors (BWRs). This system measures thehelium released from failed control rods containing boron carbide (B4C). However,there are a number of factors that might influence measurements, which have to betaken into consideration when evaluating the measured data. These factors can beseparated into two groups: 1) local adjustments, made on the sampling line connectingthe detector to the off-gas system, and 2) plant operational parameters. The adjustmentsof the sample line conditions include variation of gas flow rate and gas pressurein the line. Plant operational factors that may influence helium measurements canvary from plant to plant. The factors studied at Leibstadt nuclear power plant (KKL)were helium impurities in injected hydrogen gas, variation of the total off-gas flow andregular water refill. In this paper we discuss these factors and their significance andpresent experimental results of measurements at KKL.8:25 AMAssessment of the Mechanical Performance of the WestinghouseBWR Control Rod CR 99 at High Depletion LevelsPer Seltborg and Magnus JinnestrandWestinghouse Electric Sweden AB, Västerås, SwedenA long-term program assessing the mechanical performance of the WestinghouseBWR control rod CR 99 at high depletion levels has been performed. The scope of theprogram has mainly been based on the operation of four CR 99 Generation 2 controlrods in demanding positions during 6 and 7 cycles in the Leibstadt Nuclear PowerPlant (KKL) and on the detailed visual inspections and blade wing thickness measurementsthat were performed after the rods were discharged. By correlating statisticallythe blade wing thickness measurements to the appearance of irradiationassistedstress corrosion cracking (IASCC), the probability of IASCC appearance as function ofthe blade wing swelling was estimated. In order to correlate the IASCC probability ofa CR 99 to its depletion, the 10B depletion of the studied rods was calculated in detailon a local level with the stochastic Monte Carlo code MCNP in combination with theWestinghouse nodal code system PHOENIX4/POLCA7. Using this information coupledto the blade wing measurement data, a finite element model describing the bladewing swelling of an arbitrary CR 99 design as function of 10B depletion could then begenerated. In the final step, these relationships were used to quantify the probability ofIASCC appearance as function of the 10B depletion of the CR 99 Generations 2 and3. Applying this detailed mapping of the CR 99 behavior at high depletion levels andusing an online core monitoring system with explicit 10B depletion tracking capabilitieswill enable a reliable prediction of the probability for IASCC appearance, thus enhancingthe optimized design and the sound operation of the CR 99 control rod. Anotherimportant outcome of the program was that it was clearly shown that no significantamount of boron leakage did occur through any of the detected IASCC cracks, despitethe very high depletion levels achieved.8:50 AMFuel Cycle Cost, Reactor Physics and Fuel ManufacturingConsiderations for Erbia-Bearing PWR Fuel with > 5Wt%U-235 ContentFausto Franceschini, Edward J. Lahoda and Vefa N. KucukboyaciWestinghouse Electric Co. LLC, Cranberry Township, PAThe efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limitin fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on currentZr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, whilecertainly possible, entails costly investment in infrastructure and licensing. As a possibleway to offset some of these costs, the addition of small amounts of erbia to theUO2 powder with >5 wt% U-235 has been proposed, so that its initial reactivity isreduced to that of licensed fuel and most modifications to the existing facilities andequipment could be avoided. This paper discusses the potentialities of such a fuel onthe US market from a vendor’s perspective. An analysis of the in-core behavior andfuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycleshas been conducted, with the aim of quantifying the potential economic advantageand other operational benefits of this concept. Subsequently, the implications on fuelmanufacturing and storage are discussed. While this concept has certainly good potential,a compelling case for its short-term introduction as PWR fuel for the US marketcould not be determined.9:15 AMEffect of Twinning on Texture Evolution of Depleted UraniumUsing a Viscoplastic Self-Consistent ModelJohn Ho and Hamid Garmestani (1), Reeshemah Burrell and Anthony Belvin(2), Dongsheng Li (3), David McDowell (4), Anthony Rollett (5)1) Georgia Institute of Technology, Atlanta, GA, USA. 2) Y-12 National Security ComplexOak Ridge, TN. 3) Fundamental and Computational Sciences DirectoratePacific Northwest National Laboratory, Richland, WA . 4) Woodruff School of Mechanical Engineering,Atlanta, GA. 5) Department of Materials Science and EngineeringCarnegie Mellon University, Pittsburgh, PADuctility and fracture toughness is a major stumbling block in using depleted uraniumas a structural material. The ability to correctly model deformation of uranium can beused to create process path methods to improve its structural design ability. The texturalevolution of depleted uranium was simulated using a visco-plastic self consistentmodel and analyzed by comparing pole figures of the simulations and experimentalsamples. Depleted uranium has the same structure as alpha uranium, which is anorthorhombic phase of uranium. Both deformation slip and twin systems were compared.The VPSC model was chosen to simulate this material because the modelencompasses both low-symmetry materials as well as twinning in materials. This isof particular interest since depleted uranium has a high propensity for twinning, whichdominates deformation and texture evolution. Simulated results were compared toexperimental results to measure the validity of the model. One specific twin system,the {176}[512] twin, was of specific notice. The VPSC model was used to simulatethe influence of this twin on depleted uranium and was compared with a mechanicallyshocked depleted uranium sample. Under high strain rate shock deformation conditions,the {176}[512] twin system appears to be a dominant deformation system. Bysimulating a compression process using the VPSC model with the {176}[512] twin asthe dominant deformation mode, a favorable comparison could be made between theexperimental and simulated textures.9:40 AMThe Relative Variational Model - A Topological View of Matterand its Properties: Space Occupancy by the AtomsMarcio S. Dias, Vanderley de Vasconcelos, João Roberto L. Mattos (1),Elizabete Jordão (2)1) Center for Development of the Nuclear Technology – CDTN, National Commission for the NuclearEnergy – CNEN, Minas Gerais, Brazil. 2) Chemistry Engineering Dept., Campinas State University, FEQ/UNICAMP, São Paulo, BrazilFormal definitions of convergence, connectedness and continuity were establishedto characterize and describe the crystalline solid and its properties as a unified notionin the topological space. In this unified notion, physical and material properties aremodeled by means of an intrinsic and invariable form function: the Relative VariationalModel. The crystalline solid is assumed an empty space that has been filled with atomsand phonons, i.e., the crystal is built with packages of matter and energy in a regularand orderly repetitive pattern along three orthogonal dimensions of the space. Thespatial occupation of the atom in the crystalline structure is determined by its meanvibrational volume, which also defines the lattice parameter or interatomic distance.However, as packages of vibrational energy, phonons can only exist as vibrations ofatoms. Any variation of internal energy is in fact the discretized variations of phonon’spopulation. These variations occur in the quantized modes of vibration, and thereforethe balance between the frequency and amplitude of vibrations also is a dynamic variable.In this paper, the Relative Variational Model was applied to deconvolutions offrequency spectra of the inelastic neutron scatterings. Some dynamic aspects of atomvibration were presented and evaluated in support to the model’s fundamentals.26

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 8:00 AM - 301 B2C - Deterministic Transport TheorySession Chair: Bob Grove (ORNL); Piero Ravetto (PdiT)8:00 AMRayleigh Quotient Iteration in 3D, Deterministic NeutronTransportR. N. Slaybaugh (1), T. M. Evans and G. G. Davidson (2)1) Department of Nuclear Engineering and Engineering Physics, University of Wisconsin - Madison,Madison, WI. 2) Radiation Transport Group, Oak Ridge National Laboratory, Oak Ridge, TNToday’s “grand challenge” neutron transport problems require 3-D meshes with billionsof cells, hundreds of energy groups, and accurate quadratures and scatteringexpansions. Leadership-class computers provide platforms on which high-fidelity fluxescan be calculated. However, appropriate methods are needed that can use thesemachines effectively. Such methods must be able to to use hundreds of thousands ofcores and have good convergence properties. Rayleigh quotient iteration (RQI) is aneigenvalue solver that has been added to the SN code Denovo to address convergence.Rayleigh quotient iteration is an optimal shifted inverse iteration method thatshould converge in fewer iterations than the more common power method and othershifted inverse iteration methods for many problems of interest. Denovo’s RQI uses anew multigroup Krylov solver for the fixed source solutions inside every iteration thatallows parallelization in energy in addition to space and angle. This Krylov solver hasbeen shown to scale successfully to 200,000 cores: for example one test problemscaled from 69,120 cores to 190,080 cores with 98% efficiency. This paper shows thatRQI works for some small problems. However, the Krylov method upon which it reliesdoes not always converge because RQI creates ill-conditioned systems. This resultleads to the conclusion that preconditioning is needed to allow this method to be applicableto a wider variety of problems.8:25 AMLinear Source Approximation in CASMO5Rodolfo Ferrer and Joel Rhodes (1), Kord Smith (2)1) Studsvik Scandpower, Inc., Idaho Falls, ID. 2) Department of Nuclear Science and Engineering, MassachusettsInstitute of Technology, Cambridge, MAA Linear Source (LS) approximation has been implemented in the two-dimensionalMethod of Characteristics (MOC) transport solver in a prototype version of CASMO5.The LS approximation, which relies on the computation of trajectory-based spatialmoments over source regions to obtain the linear source expansion coefficients, improvesthe solution accuracy relative to the ‘flat’ or constant source approximation. Inaddition, the LS formulation is capable of treating arbitrarily-shaped source regionsand is compatible with standard Coarse-Mesh Finite Difference (CMFD) acceleration.Numerical tests presented in this paper for the C5G7 MOX benchmark show that, forcomparable accuracy with respect to the reference solution, the LS approximation canreduce the run time by a factor of four and the memory requirements by a factor of tenrelative to the FS scheme.9:15 AMNodal Collocation Approximation for The Multidimensional PLEquations Applied to Transport Source ProblemsG. Verdú (1), M. Capilla, C.F. Talavera and D. Ginestar (2)1) Departamento de Ingeniería Química y Nuclear, Universitat Politècnica de València, València, Spain.2) Department of Nuclear Engineering, Departamento de Matematica Aplicada, Universitat Politècnicade València, València, SpainPL equations are classical high order approximations to the transport equations whichare based on the expansion of the angular dependence of the angular neutron ux andthe nuclear cross sections in terms of spherical harmonics. A nodal collocation methodis used to discretize the PL equations associated with a neutron source transport problem.The performance of the method is tested solving two 1D problems with analyticalsolution for the transport equation and a classical 2D problem.9:40 AMNonlinear Acceleration of a Continuous Finite Element Discretizationof fhe Self-Adjoint Angular Flux Form of the TransportEquationRichard Sanchez (1), Cristian Rabiti and Yaki Wang (2)1) Commissariat à l’Energie Atomique et aux Energies Alternatives, Centre de Saclay, DEN/DM2S/SER-MA. 2) Idaho National Laboratory, Idaho Falls, ID, USANonlinear acceleration of a continuous finite element (CFE) discretization of the transportequation requires a modification of the transport solution in order to achieve localconservation, a condition used in nonlinear acceleration to define the stopping criterion.In this work we implement a coarse-mesh finite difference acceleration for a CFEdiscretization of the second-order self adjoint angular flux (SAAF) form of the transportequation and use a post processing to enforce local conservation. Numerical resultsare given for one-group source calculations of one-dimensional slabs. We also give aformal derivation of the boundary conditions for the SAAF.8:50 AMA 1D Analysis Of Two High Order MOC MethodsMatthew S. Everson and Benoit ForgetMassachusetts Institute of Technology, Cambridge, MAThe work presented here provides two different methods for evaluating angular fluxesalong long characteristics. One is based off a projection of the 1D transport equationonto a complete set of Legendre polynomials, while the other uses the 1D integraltransport equation to evaluate the angular flux values at specific points alongeach track passing through a cell. The Moment Long Characteristic (M-LC) method isshown to provide 2(P+1) spatial convergence and significant gains in accuracy withthe addition of only a few spatial degrees of freedom. The M-LC method, though, isshown to be ill-conditioned at very high order and for optically thin geometries. ThePoint Long Characteristic (P-LC) method, while less accurate, significantly improvesstability to problems with optically thin cells. The P-LC method is also more flexible,allowing for extra angular flux evaluations along a given track to give a more accuraterepresentation of the shape along each track. This is at the expense of increasingthe degrees of freedom of the system, though, and requires an increase in memorystorage. This work concludes that both may be used simultaneously within the samegeometry to provide the best mix of accuracy and stability possible.27

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 8:00 AM - 301 C4C - Reactor Concepts & DesignsSession Chair: Mark Porter (TVA)8:00 AMConceptual Design of a Pressure Tube Light Water Reactorwith Variable Moderator ControlReuven Rachamin and Emil Fridman (1), Alexander Galperin (2)1) Reactor Safety Division, Institute of Resource Ecology, Helmholtz-Zentrum Dresden-Rossendorf,Dresden, Germany. 2) Department of Nuclear Engineering, Ben-Gurion University of the Negev, BeerSheva, IsraelThis paper presents the development of innovative pressure tube light water reactorwith variable moderator control. The core layout is derived from a CANDU line of reactorsin general, and advanced ACR-1000 design in particular. It should be stressedhowever, that while some of the ACR-1000 mechanical design features are adopted,the core design basics of the reactor proposed here are completely different. First, theinter fuel channels spacing, surrounded by the calandria tank, contains a low pressuregas instead of heavy water moderator. Second, the fuel channel design features anadditional/external tube (designated as moderator tube) connected to a separate moderatormanagement system. The moderator management system is design to vary themoderator tube content from “dry” (gas) to “flooded” (light water filled). The dynamicvariation of the moderator is a unique and very important feature of the proposeddesign. The moderator variation allows an implementation of the “breed & burn” modeof operation. The “breed & burn” mode of operation is implemented by keeping themoderator tube empty (“dry” filled with gas) during the breed part of the fuel depletionand subsequently introducing the moderator by “flooding” the moderator tube for the“burn” part. This paper assesses the conceptual feasibility of the proposed conceptfrom a neutronics point of view.8:25 AMMYRRHA: A Multi-purpose hYbrid Research Reactor for HightechApplicationsHamid Aït Abderrahim, Peter BaetenSCK•CEN, Mol, BelgiumMYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is theflexible experimental accelerator driven system (ADS) in development at SCK•CEN.MYRRHA is able to work both in subcritical (ADS) as in critical mode. In this way,MYRRHA will allow fuel developments for innovative reactor systems, material developmentsfor generation IV (GEN IV) systems, material developments for fusion reactors,radioisotope production and industrial applications, such as Si-doping. MYRRHAwill also demonstrate the ADS full concept by coupling the three components (accelerator,spallation target and subcritical reactor) at reasonable power level to allowoperation feedback, scalable to an industrial demonstrator and allow the study of efficienttransmutation of high-level nuclear waste. MYRRHA is based on the heavy liquidmetal technology and so it will contribute to the development of lead fast reactor (LFR)technology and in critical mode, MYRRHA will play the role of European technologypilot plant in the roadmap for LFR. In this paper the historical evolution of MYRRHAand the rationale behind the design choices is presented and the latest configurationof the reactor core and primary system is described.8:50 AMA 48-Month Extended Fuel Cycle for the B&W mPower SmallModular Nuclear ReactorMadalina Aimee ErighinThe Babcock & Wilcox Company, Lynchburg, VAThe B&W mPower reactor is a small, rail-shippable pressurized water reactor(PWR) with an integral once-through steam generator and an electric power output of150 MW, which is intended to replace aging fossil power plants of similar output. Thecore is composed of 69 reduced-height, but otherwise standard, PWR assemblies withthe familiar 17x17 fuel rod array on a 21.5 cm inter-assembly pitch. The B&W mPowercore design and cycle management plan, which were performed using the Studsvikcore design code suite, follow the pattern of a typical nuclear reactor fuel cycle designand analysis performed by most nuclear fuel management organizations, such as fuelvendors and utilities. However, B&W is offering a core loading and cycle managementplan for four years of continuous power operations without refueling and without thehurdles of chemical shim.9:15 AMBWR Fuel Design Options for Self-Sustainable Th-233U FuelCycleYaniv Shaposhnik and Eugene Shwageraus (1), Ezra Elias (2)1) Ben-Gurion University of the Negev, Department of Nuclear Engineering, Beer-Sheva, Israel. 2) Facultyof Mechanical Engineering, Technion - Israel Institute of Technology, Haifa, IsraelIn this work, we investigate a number of fuel assembly design options for a BWR coreoperating in a closed self-sustainable Th-233U fuel cycle. The designs rely on axiallyheterogeneous fuel assembly structure in order to improve fertile to fissile conversionratio. One of the main assumptions of the current study was to restrict the fuel assemblygeometry to a single axial fissile zone “sandwiched” between two fertile blanketzones. The main objective was to study the effect of the most important design parameters,such as dimensions of fissile and fertile zones and average void fraction, on thenet breeding of 233U. The main design challenge in this respect is that the fuel breedingpotential is at odds with axial power peaking and therefore limits the maximumachievable core power rating. The calculations were performed with BGCore system,which consists of MCNP code coupled with fuel depletion and thermo-hydraulic feedbackmodules. A single 3-dimensional fuel assembly with reflective radial boundarieswas modeled applying simplified restrictions on maximum central line fuel temperatureand Critical Power Ratio. It was found that axially heterogeneous fuel assembly designwith single fissile zone can potentially achieve net breeding. In this case however, theachievable core power density is roughly one third of the reference BWR core.9:40 AMComputation of a Canadian SCWR Unit Cell with Deterministicand Monte Carlo CodesG. Harrisson and G. MarleauInstitute of Nuclear Engineering, École Polytechnique de Montréal, Montréal, QuébecThe Canadian SCWR has the potential to achieve the goals that the generation IVnuclear reactors must meet. As part of the optimization process for this design concept,lattice cell calculations are routinely performed using deterministic codes. In thisstudy, the first step (self-shielding treatment) of the computation scheme developedwith the deterministic code DRAGON for the Canadian SCWR has been validated.Some options available in the module responsible for the resonance self-shieldingcalculation in DRAGON 3.06 and different microscopic cross section libraries basedon the ENDF/B-VII.0 evaluated nuclear data file have been tested and compared toa reference calculation performed with the Monte Carlo code SERPENT under thesame conditions. Compared to SERPENT, DRAGON underestimates the infinite multiplicationfactor in all cases. In general, the original Stammler model with the Livolant-Jeanpierre approximations are the most appropriate self-shielding options to use inthis case of study. In addition, the 89 groups WIMS-AECL library for slight enricheduranium and the 172 groups WLUP library for a mixture of plutonium and thorium givethe most consistent results with those of SERPENT.28

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 8:00 AM - 301 DSS2C - Radiation Transport Methods for Whole Reactor Core AnalysisSession Chair: F. Rahnema (GT); Tom Sutton (KAPL)8:00 AMDevelopment of a Neutronics Calculation Method for DesigningCommercial Type Japanese Sodium-Cooled Fast ReactorToshikazu Takeda, Yoichiro Shimazu, Koki Hibi, and Koji FujimuraResearch Institute of Nuclear Engineering, University of Fukui, Fukui, JapanUnder the R&D project to improve the modeling accuracy for the design of fast breederreactors the authors are developing a neutronics calculation method for designing alarge commercial type sodium- cooled fast reactor. The calculation method is establishedby taking into account the special features of the reactor such as the use ofannular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verificationand Validation, and Uncertainty Qualification (V&V&UQ) of the calculation method isbeing performed by using measured data from the prototype FBR Monju. The resultsof this project will be used in the design and analysis of the commercial type demonstrationFBR, known as the Japanese Sodium fast Reactor (JSFR).8:25 AMHybrid Parallel Code Acceleration Methods in Full-Core ReactorPhysics CalculationsTanguy Courau, Laurent Plagne, and Angélique Ponçot (1), Glenn Sjoden(2)1) EDF R&D, Clamart Cedex, France. 2) Nuclear and Radiological Engineering, Georgia Institute ofTechnology, Atlanta, GAWhen dealing with nuclear reactor calculation schemes, the need for three dimensional(3D) transport-based reference solutions is essential for both validation andoptimization purposes. Considering a benchmark problem, this work investigates thepotential of discrete ordinates (Sn) transport methods applied to 3D pressurized waterreactor (PWR) full-core calculations. First, the benchmark problem is described. Itinvolves a pin-by-pin description of a 3D PWR first core, and uses a 8-group crosssectionlibrary prepared with the DRAGON cell code. Then, a convergence analysisis performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatialrefinement and the associated angular quadrature required to properly describe theproblem physics. It also shows that initializing the Sn solution with the EDF SPN solverCOCAGNE reduces the number of iterations required to converge by nearly a factorof 6. Using a best estimate model, PENTRAN results are then compared to multigroupMonte Carlo results obtained with the MCNP5 code. Good consistency is observedbetween the two methods (Sn and Monte Carlo), with discrepancies that are less than25 pcm for the ke , and less than 2.1% and 1.6% for the flux at the pin-cell level andfor the pin-power distribution, respectively.8:50 AMA Hybrid Monte Carlo and Response Matrix Monte CarloMethod in Criticality CalculationZeguang Li And Kan WangDepartment of Engineering Physics, Tsinghua University, Beijing, ChinaFull core calculations are very useful and important in reactor physics analysis, especiallyin computing the full core power distributions, optimizing the refueling strategiesand analyzing the depletion of fuels. To reduce the computing time and acceleratethe convergence, a method named Response Matrix Monte Carlo (RMMC) methodbased on analog Monte Carlo simulation was used to calculate the fixed source neutrontransport problems in repeated structures. To make more accurate calculations,we put forward the RMMC method based on nonanalog Monte Carlo simulation andinvestigate the way to use RMMC method in criticality calculations. Then a new hybridRMMC and MC (RMMC+MC) method is put forward to solve the criticality problemswith combined repeated and flexible geometries. This new RMMC+MC method, havingthe advantages of both MC method and RMMC method, can not only increase theefficiency of calculations, also simulate more complex geometries rather than repeatedstructures. Several 1-D numerical problems are constructed to test the new RMMCand RMMC+MC method. The results show that RMMC method and RMMC+MC methodcan efficiently reduce the computing time and variations in the calculations. Finally,the future research directions are mentioned and discussed at the end of this paper tomake RMMC method and RMMC+MC method more powerful.9:15 AMA Multi-Group Monte Carlo Core Analysis Method and its Applicationin SCWR DesignZhang Peng, Wang Kan, Yu GanglinDepartment of Engineering Physics, Tsinghua University, Beijing, ChinaComplex geometry and spectrum have been the characteristics of many newly developednuclear energy systems, so the suitability and precision of the traditional deterministiccodes are doubtable while being applied to simulate these systems. Onthe contrary, the Monte Carlo method has the inherent advantages of dealing withcomplex geometry and spectrum. The main disadvantage of Monte Carlo method isthat it takes long time to get reliable results, so the efficiency is too low for the ordinarycore designs. A new Monte Carlo core analysis scheme is developed, aimed toincrease the calculation efficiency. It is finished in two steps: Firstly, the assembly levelsimulation is performed by continuous energy Monte Carlo method, which is suitablefor any geometry and spectrum configuration, and the assembly multi-group constantsare tallied at the same time; Secondly, the core level calculation is performed by multigroupMonte Carlo method, using the assembly group constants generated in the firststep. Compared with the heterogeneous Monte Carlo calculations of the whole core,this two-step scheme is more efficient, and the precision is acceptable for the preliminaryanalysis of novel nuclear systems. Using this core analysis scheme, a SCWRcore was designed based on a new SCWR assembly design. The core output is about1,100 MWe, and a cycle length of about 550 EFPDs can be achieved with 3-batchrefueling pattern. The average and maximum discharge burn-up are about 53.5 and60.9 MWD/kgU respectively.9:40 AMABWR Start-Up Test Analysis Using BWR Core Simulator withThree-Dimensional Direct Response Matrix MethodTakeshi Mitsuyasu, Kazuya Ishii, Tetsushi Hino and Motoo AoyamaHitachi, Ltd., Hitachi Research Laboratory, 7-chome Hitachi-shi Ibaraki-ken, JapanThe ABWR start-up test analysis has been done with the BWR core simulator usingthe three-dimensional direct response matrix (3D-DRM) method. The Monte Carlocode VMONT made the sub-response matrices for the 3D-DRM method. Each boundarysurface was subdivided by 4 × 4 for transverse segments, by 4 for angular segmentsand by 4 for axial zones in a node. For the calculation speedup, the 3D-DRMcode used the divided sub-response matrices data set. The code used the MPI andOpenMP for the parallelized method. The median value is set as the average criticaleigenvalues. The changes from the maximum value to the minimum value are0.34 %Δk with the spectral history method and 0.40 %Δk without it, and the respectivestandard deviations were 0.12 % and 0.14 %. Using the spectral history methoddecreased the variation by 0.06 %Δk. The root mean square differences of the axialpower distribution were about 6 % between the analysis results and the plant data. Usingthe currents which converged in the previous exposure step reduced the numberof iterations when the CR pattern changed only slightly. The averaged calculation timefor each exposure step was about 5 hours on 12 PC Linux cluster servers with Core2 Quad 3GHz.29

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 8:00 AM - 301 E8B - Advanced Modeling & Simulation in Reactor PhysicsSession Chair: Bill Martin (U. Mich); Steve Bowman (ORNL)8:00 AMA Multidimensional and Multiphysics Approach To NuclearFuel Behavior SimulationS. R. Novascone, R. L. Williamson, J. D. Hales, M. R. Tonks, D. R. Gaston,C. J. Permann, D. Andrs, R. C. MartineauIdaho National Laboratory, Idaho Falls, ID USAImportant aspects of fuel rod behavior are inherently multidimensional in addition tobeing complicated multiphysics problems. However, many current fuels modelingtools are strictly 2D axisymmetric or even 1.5D. This paper outlines the capabilities ofa new fuel modeling tool able to analyze either 2D axisymmetric or fully 3D models.The need for multiphysics, multidimensional modeling is then demonstrated through aset of example problems. The first, a 10-pellet rodlet, demonstrates the viability of thesolution method employed. This example highlights the effect of our smeared crackingmodel and also shows the multidimensional nature of discrete fuel pellet modeling.The second example relies on our multidimensional, multiphysics approach to analyzea missing pellet surface problem. As a final example, we show a 5-pellet rodlet simulationcoupled to lower-length-scale simulations of bubble growth.8:25 AMA Coupled TH/Neutronics/CRUD Framework in Prediction ofCIPS PhenomenonLing Zou and Hongbin Zhang (1), Jess Gehin (2), Brendan Kochunas (3)1) Idaho National Laboratory, Idaho Falls, ID. 2) Oak Ridge National Laboratory, Oak Ridge, TN. 3) Departmentof Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MIA coupled TH/Neutronics/CRUD framework, which is able to simulate the CRUD depositsimpact on CIPS phenomenon, was described in this paper. The coupling amongthree essential physics, thermal-hydraulics, CRUD and neutronics described in theframework was implemented by using CFD software STAR-CCM+, developing CRUDmodule, and using the neutronics code DeCART. The coupling among these codeswas implemented by exchanging data between them using intermediate exchangefiles. A typical 3 by 3 PWR fuel pin problem was solved under this framework and theresults were presented. Time-dependent solutions were provided for a 12- month simulation,including CRUD deposits inventory and their distributions on fuel rods, boronhideout amount inside CRUD deposits, as well as power shape changing over time.The results clearly showed the power shape suppression in regions where CRUDdeposits exist, which is a clear indication of CIPS phenomenon.8:50 AMCoupled Thermal Analysis Applied to the Study of the RodEjection AccidentMichel GonnetAREVA NP, TOUR AREVA, PARIS LA DÉFENSE CEDEX, FranceAn advanced methodology for the assessment of fuel-rod thermal margins underRIA conditions has been developed by AREVA NP SAS. With the emergence of RIAanalytical criteria, the study of the Rod Ejection Accident (REA) would normally requirethe analysis of each fuel rod, slice by slice, over the whole core. Up to now thestrategy used to overcome this difficulty has been to perform separate analyses ofsampled fuel pins with conservative hypotheses for thermal properties and boundaryconditions. In the advanced methodology, the evaluation model for the Rod EjectionAccident (REA) integrates the node average fuel and coolant properties calculationfor neutron feedback purpose as well as the peak fuel and coolant time-dependentproperties for criteria checking. The calculation grid for peak fuel and coolant propertiescan be specified from the assembly pitch down to the cell pitch. The comparativeanalysis of methodologies shows that coupled methodology allows reducing excessiveconservatism of the uncoupled approach.9:15 AMModel Biases in High-Burnup Fast Reactor SimulationsNicholas Touran, Jesse Cheatham, and Robert PetroskiTerraPower LLC, Bellevue, WAA new code system called the Advanced Reactor Modeling Interface (ARMI) has beendeveloped that loosely couples multiscale, multiphysics nuclear reactor simulations toprovide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic,thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economicanalyses, ARMI provides “one-click” assessments of many multi-disciplined performancemetrics and constraints that historically require iterations between many diverseexperts. The capabilities of ARMI are implemented in this study to quantify neutronicbiases of various modeling approximations typically made in fast reactor analysis atan equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium thatresult in very high discharge burnup are considered (>20% FIMA), as motivated by thedevelopment of the Traveling Wave Reactor. Model approximations discussed includehomogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling,explicit control rod insertion, burnup-dependent cross sections, fission product model,burn chain truncation, and dynamic fuel performance. The sensitivities of these approximationson equilibrium discharge burnup, keff, power density, delayed neutronfraction, and coolant temperature coefficient are discussed.9:40 AMModeling Prismatic HTGRs With U.S. N.R.C Advanced GasReactor Evaluator (AGREE)Volkan Seker, Timothy Drzewiecki, Thomas Downar (1), Joseph M. Kelly(2)1) Nuclear Engineering and Radiological Sciences, Ann Arbor, MI. 2) US Nuclear Regulatory Commission,Washington, D.C. USAA core fluids and heat transfer model has been developed for the prismatic high temperaturegas reactor in support of the US NRC Next Generation Nuclear Plant (NGNP)evaluation model. The core fluids modeling relies on a subchannel approach in whichthe primary coolant flowpath through the core region and vertical in-core and ex-coregaps can be modeled as individual subchannels. These subchannels are connectedtogether to represent a three dimensional reactor. An initial validation calculation forthe core fluids model has been performed using data available in literature for bypassflow. The predicted bypass flow was within 2.6% of the value reported in the literature.The core level heat transfer model is based on a triangular finite volume method,where the base triangle is one sixth of the prismatic block. In order to improve thespatial accuracy at this level, a triangular refinement method was also implemented.The fuel compact temperature is calculated by a cylindrical conduction model whichis implicitly coupled to the triangular core level model. The preliminary verification ofthe model was performed by comparing AGREE to a finite element code COMSOL byanalyzing the MHTGR core heat transfer. Further verification and validation is currentlyan ongoing effort.30

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 10:20 AM - 301 A1C - Core Analysis MethodsSession Chair: W. Zwermann (GRS)10:20 AMPhysics Model of a Gas-Cooled Fast Reactor: Review and AssessmentHangbok ChoiGeneral Atomics, San Diego, CA, U.S.A.11:10 AMDevelopment of HELIOS/CAPP Code System for the Analysisof Block Type VHTR CoresHyun Chul Lee, Tae Young Han, Chang Keun Jo, Jae Man NohKorea Atomic Energy Research Institute, Daejeon, KoreaThe current physics design and analysis model was reviewed and assessed for itsapplication to a long-life gas-cooled fast reactor (GFR) design. The physics designuses MICROX, BURP, and DIF3D for the cross section generation, depletion calculation,and criticality and flux calculation, respectively. For the application to the long-lifeGFR, the depletion model was adjusted such that more lumped fission products areincluded in the burn-up chain to preserve the reaction rate and fuel mass. The performanceof the physics design tools including the adjustment of the depletion model wasassessed against Monte Carlo depletion calculations. The comparison has shown thatthe excess reactivity and cycle length of the long-life GFR are reasonably predicted.Some discrepancies were found at the beginning of cycle, which can be attributed tothe differences between the nuclear data used in each model. Further studies will becarried out to update the cross section library of the MICROX code for agreement withthe latest sets and to expand the fuel burn-up chain for the high burn-up and recyclingfuel cycle analysis.10:45 AMStochastic Sampling Method with MCNPX for Nuclear DataUncertainty Propagation in Criticality Safety ApplicationsT. Zhu, A. Vasiliev, W. Wieselquist and H. FerroukhiPaul Scherrer Institut, Villigen PSI,SwitzerlandIn the domain of criticality safety, the efficient propagation of uncertainty in nucleardata to uncertainty in keff is an important area of current research. In this paper, amethod based on stochastic sampling is presented for uncertainty propagation in MC-NPX calculations. To that aim, the nuclear data (i.e. cross sections) are assumed tohave a multivariate normal distribution and simple random sampling is performed followingthis presumed probability distribution. A verification of the developed stochasticsampling procedure with MCNPX is then conducted using the 239Pu Jezebel experimentas well as the PB-2 BWR and TMI-1 PWR pin cell models from the UncertaintyAnalysis in Modeling (UAM) exercises. For the Jezebel case, it is found that the developedstochastic sampling approach predicts similar keff uncertainties compared toconventional sensitivity and uncertainty methods. For the UAM models, slightly loweruncertainties are obtained when comparing to existing preliminary results. Furtherdetails of these verification studies are discussed and directions for future work areoutlined.In this paper, the HELIOS/CAPP code system developed for the analysis of block typeVHTR cores is presented and verified against several VHTR core configurations. Verificationresults showes that HELIOS code predicts less negative MTC and RTC thanMcCARD code does and thus HELIOS code overestimates the multiplication factors atthe states with high moderator and reflector temperature especially when the B4C BPis loaded. In the depletion calculation for the VHTR single cell fuel element, the errorof HELIOS code increases as burnup does. It is ascribed to the fact that HELIOS codetreats some fission product nuclides with large resonances as nonresonant nuclides.In the 2-D core depletion calculation, a relatively large reactivity error is observed inthe case with BP loading while the reactivity error in the case without BP loading isless than 300 pcm.11:35 AMVerification of a Depletion Method in Scale for the AdvancedHigh Temperature ReactorRyan Kelly (1), Dan Ilas (2)1) Department of Nuclear Engineering, Texas A&M University, College Station, TX. 2) Oak Ridge NationalLaboratory, Oak Ridge, TNThis study describes a new approach employing the Dancoff correction method tomodel the TRISO-based fuel form used by the Advanced High-Temperature Reactor(AHTR) reactor design concept. The Dancoff correction method is used to performisotope depletion analysis using the TRITON sequence of SCALE and is verified bycode-to-code comparisons. The current AHTR fuel design has TRISO particles concentratedalong the edges of a slab fuel element. This geometry prevented the use ofthe DOUBLEHET treatment, previously developed in SCALE to model spherical andcylindrical fuel. The new method permits fuel depletion on complicated geometries thattraditionally can be handled only by continuous energy based depletion code systems.The method was initially tested on a fuel configuration typical of the Next GenerationNuclear Plant (NGNP), where DOUBLEHET treatment is possible. A confirmatorystudy was performed on the AHTR reference core geometry using the VESTA code,which uses the continuous energy MCNP5 code as a transport solver and ORIGEN2.2code for depletion calculations. Comparisons of the results indicate good agreement ofwhole core characteristics, such as the multiplication factor and the isotopics, includingtheir spatial distribution. Key isotopes analyzed included 235U, 239Pu, 240Pu, and241Pu. The results from this study indicate that the Dancoff factor method can generateestimates of core characteristics with reasonable precision for scoping studies ofconfigurations where DOUBLEHET treatment cannot be performed.31

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 10:20 AM - 301 B2D - Deterministic Transport TheorySession Chair: Bob Grove (ORNL)10:20 AMThe Principal Component Analysis Method Used with PolynomialChaos Expansion to Propagate Uncertainties ThroughCritical Transport ProblemsMichael E. Rising and Anil K. PrinjaUniversity of New Mexico, Department of Chemical and Nuclear Engineering, Albuquerque, NM, USA11:10 AMComparison of Homogenized and Enhanced Diffusion Solutionsof Model PWR ProblemsE. E. Lewis (1), Micheal A. Smith (2)1) Department of Mechanical Engineering, Northwestern University, Evanston, IL, USA. 2) Nuclear EngineeringDivision, Argonne National Laboratory, Argonne, IL, USAA critical neutron transport problem with random material properties is introduced. Thetotal cross section and the average neutron multiplicity are assumed to be uncertain,characterized by the mean and variance with a log-normal distribution. The averageneutron multiplicity and the total cross section are assumed to be uncorrelated and thematerial properties for differing materials are also assumed to be uncorrelated. Theprincipal component analysis method is used to decompose the covariance matrix intoeigenvalues and eigenvectors and then “realizations” of the material properties can becomputed. A simple Monte Carlo brute force sampling of the decomposed covariancematrix is employed to obtain a benchmark result for each test problem. In order to savecomputational time and to characterize the moments and probability density functionof the multiplication factor the polynomial chaos expansion method is employed alongwith the stochastic collocation method. A Gauss-Hermite quadrature set is convolvedinto a multidimensional tensor product quadrature set and is successfully used to computethe polynomial chaos expansion coefficients of the multiplication factor. Finally,for a particular critical fuel pin assembly the appropriate number of random variablesand polynomial expansion order are investigated.10:45 AMOn the Approximate Albedo Boundary Conditions for Two-Energy Group X,Y-Geometry Discrete Ordinates EigenvalueProblemsCarlos Eduardo A. Nunes, Hermes Alves Filho and Ricardo C. BarrosPrograma de Pós-graduação em Modelagem Computacional, Instituto Politécnico, IPRJ, Universidadedo Estado do Rio de Janeiro, UERJ, Nova Friburgo, RJ, BrazilWe discuss in this paper the computational efficiency of approximate discrete ordinates(SN) albedo boundary conditions for two-energy group eigenvalue problems inX,Y-geometry. The non-standard SN albedo substitutes approximately the reflectorsystem around the active domain, as we neglect the transverse leakage terms withinthe non-multiplying reflector region. Should the problem have no transverse leakageterms, i.e., one-dimensional slab geometry, then the offered albedo boundary conditionsare exact. By computational efficiency we mean analyzing the accuracy of thenumerical results versus the CPU execution time of each run for a given model problem.Numerical results to a typical test problem are shown to illustrate this efficiencyanalysis.Model problem comparisons in slab geometry are made between two forms of homogenizeddiffusion theory and enhanced diffusion theory. The pin-cell discontinuity factorsfor homogenized diffusion calculations are derived from homogenized variationalnodal P1 response matrices and from standard finite differencing. Enhanced diffusiontheory consists of applying quasi-reflected interface conditions to reduce variationalnodal Pn response matrices to one degree of freedom per interface, without homogenizationwithin the cell. As expected both homogenized diffusion methods preservereaction rates exactly if the discontinuity factors are derived from the P11 referencesolutions. If no reference lattice solution is available, discontinuity factors may be approximatedfrom single cells with reflected boundary conditions; the computationaleffort is then comparable to calculating the enhanced diffusion response matrices. Inthis situation enhanced diffusion theory gives the most accurate results and finite differencediscontinuity factors the least accurate.11:35 AMCMFD Acceleration of Spatial Domain-Decomposed NeutronTransport ProblemsBlake W. Kelley and Edward W. LarsenDepartment of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MichiganUSAA significant limitation to parallelizing the solution of neutron transport problems is theneed for sweeps across the entirety of the problem domain. Angular domain decompositionis common practice, as the equations for each direction are independent asidefrom their shared scattering/fission source. Accordingly, spatial domain decompositiondoes not naturally arise in the transport equations and is therefore less frequent inpractice. In this paper, we show that a neutron transport domain can be straightforwardlydivided into independent, parallelizable sweep regions, globally linked with thestandard CMFD method, with an additional update equation. We verify, theoretically(via Fourier analysis) and computationally, that the convergence properties of thismethod are stable and nominally as rapid as standard CMFD.32

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 10:20 AM - 301 C3B - Monte Carlo Methods & DevelopmentsSession Chair: John Wagner (ORNL); Tom Sutton (KAPL)10:20 AMAcceleration of Monte Carlo Criticality Calculations UsingDeterministic-Based Starting SourcesAhmad M. Ibrahim (1), Douglas E. Peplow, John C. Wagner, Scott W. Mosher,and Thomas M. Evans (2)1) University of Wisconsin-Madison, Madison, WI. 2) Oak Ridge National Laboratory, Oak Ridge, TNA new automatic approach that uses approximate deterministic solutions for providingthe starting fission source for Monte Carlo eigenvalue calculations was evaluated inthis analysis. By accelerating the Monte Carlo source convergence and decreasing thenumber of cycles that has to be skipped before the tallies’ estimation, this approachwas found to increase the efficiency of the overall simulation, even with the inclusion ofthe extra computational time required by the deterministic calculation. This approachwas also found to increase the reliability of the Monte Carlo criticality calculations ofloosely coupled systems because the use of the better starting source reduces thelikelihood of producing an undersampled keff due to the inadequate source convergence.The efficiency improvement was demonstrated using two of the standard testproblems devised by the OECD/NEA Expert Group on Source Convergence in Criticality-SafetyAnalysis to measure the source convergence in Monte Carlo criticalitycalculations. For a fixed uncertainty objective, this approach increased the efficiencyof the overall simulation by factors between 1.2 and 3 depending on the difficulty of thesource convergence in these problems. The reliability improvement was demonstratedin a modified version of the “keff of the world” problem that was specifically designed todemonstrate the limitations of the current Monte Carlo power iteration techniques. Forthis problem, the probability of obtaining a clearly undersampled keff decreased from5% with a uniform starting source to zero with a deterministic starting source whenbatch sizes with more than 15,000 neutron/cycle were used.10:45 AMMonte Carlo Reactor Calculation with Substantially ReducedNumber of CyclesMin Jae Lee and Han Gyu Joo (1), Deokjung Lee (2), Kord Smith (3)1) Seoul National University, Gwanak-gu, Seoul, Korea. 2) Ulsan National Institute of Science and Technology,Ulsan, Korea. 3) Massachusetts Institute of Technology, Cambridge, MA, USAA new Monte Carlo (MC) eigenvalue calculation scheme that substantially reduces thenumber of cycles is introduced with the aid of coarse mesh finite difference (CMFD)formulation. First, it is confirmed in terms of pin power errors that using extremelymany particles resulting in short active cycles is beneficial even in the conventional MCscheme although wasted operations in inactive cycles cannot be reduced with moreparticles. A CMFD-assisted MC scheme is introduced as an effort to reduce the numberof inactive cycles and the fast convergence behavior and reduced inter-cycle effectof the CMFD assisted MC calculation is investigated in detail. As a practical meansof providing a good initial fission source distribution, an assembly based few-groupcondensation and homogenization scheme is introduced and it is shown that efficientMC eigenvalue calculations with fewer than 20 total cycles (including inactive cycles)are possible for large power reactor problems.11:10 AMOptimization of Monte Carlo Transport Simulations in StochasticMediaChao Liang and Wei JiDepartment of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy,NYThis paper presents an accurate and efficient approach to optimize radiation transportsimulations in a stochastic medium of high heterogeneity, like the Very High TemperatureGas-cooled Reactor (VHTR) configurations packed with TRISO fuel particles.Based on a fast nearest neighbor search algorithm, a modified fast Random SequentialAddition (RSA) method is first developed to speed up the generation of the stochasticmedia systems packed with both mono-sized and poly-sized spheres. A fast neutrontracking method is then developed to optimize the next sphere boundary search in theradiation transport procedure. In order to investigate their accuracy and efficiency, thedeveloped sphere packing and neutron tracking methods are implemented into an inhousecontinuous energy Monte Carlo code to solve an eigenvalue problem in VHTRunit cells. Comparison with the MCNP benchmark calculations for the same problemindicates that the new methods show considerably higher computational efficiency.11:35 AMVariance Estimation in Domain Decomposed Monte CarloEigenvalue CalculationsBrenden T. Mervin (1), Scott W. Mosher, Thomas M. Evans, John C. Wagner(2), and G I. Maldonado (1)1) Department of Nuclear Engineering, University of Tennessee, Knoxville, TN. 2) Oak Ridge NationalLaboratory, Oak Ridge, TNThe number of tallies performed in a given Monte Carlo calculation is limited in mostmodern Monte Carlo codes by the amount of memory that can be allocated on a singleprocessor. By using domain decomposition, significantly more tallies can be performedbecause the calculation is now limited by the total amount of memory available on allprocessors. Unfortunately, decomposing the problem geometry also introduces significantissues with the way tally statistics are conventionally calculated. In order to dealwith the issue of calculating tally variances in domain decomposed environments forthe Shift hybrid Monte Carlo code, this paper presents an alternative approach for reactorscenarios in which an assumption is made that once a particle leaves a domain,it does not reenter the domain. Particles that reenter the domain are instead treatedas separate independent histories. This assumption introduces a bias that generallyleads to under-prediction of the calculated variances for tallies within a few mean freepaths of the domain boundaries. However, through the use of different decompositionstrategies, primarily overlapping domains, the under-prediction in the calculateduncertainties is significantly reduced. In the cases considered in these analyses, andusing domains with an overlap fraction of 0.5, the estimated error in the standarddeviation for all tally cells is within 20%, within 10% for over 99% of the tally cells, andwithin 5% for over 75% of the tally cells.33

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 10:20 AM - 301 D4D - Reactor Concepts & DesignsSession Chair: Mark Porter (TVA)10:20 AMEvaluation of the Adequacy of Using FEW-Group Lattice- HomogenizedProperties for the Diffusion Analysis Of The SuperCritical Water ReactorWei ShenCandu Energy Inc., Mississauga, ON, CanadaTwo issues may affect the accuracy of computed core reactivities and flux/powerdistributions for the Super Critical Water Reactor (SCWR) core with traditional coreanalysiscode RFSP: one is the two-energy-group neutron-diffusion theory; the otheris the generation of lattice-homogenized properties with the lattice code based on thesingle-lattice-cell model without considering the effects of the environment. These twoissues are not SCWR specific; however their effect may be more significant for SCWR.It has been illustrated that the lattice-homogenized properties calculated with the single-lattice-cellmodel is not sufficiently accurate for heterogeneous core configurationssuch as ACR-1000 checkerboard-voiding and core-reflector interface when adjacentchannels experience significant spectrum interaction. To evaluate the adequacy ofusing two-group neutron-diffusion theory with single-lattice-based lattice properties forthe analysis of the SCWR core, a 2-D SCWR benchmark problem was setup with thereference solution provided by the continuous-energy Monte-Carlo code SERPENT.The assessment shows that the traditional two-group neutron-diffusion theory with thesingle-lattice-cell-based lattice properties is not sufficient to capture either the spectralchange or the environment effect for the SCWR core. The solution of the eight-groupneutron-diffusion equation with the multicell- based lattice properties is consideredappropriate for the analysis of the d SCWR core.10:45 AMCore Design Study of a Supercritical Light Water ReactorWith Double Row Fuel RodsChuanqi Zhao, Hongchun Wu, Liangzhi Cao, Youqi Zheng (1), Jue Yang,Yong Zhang (2)1) School of nuclear Science and Technology, Xi’an Jiaotong University, , ShannXi, P.R. China. 2) ChinaNuclear Power Technology Research Institute, GuangDong, P.R. ChinaAn equilibrium core for supercritical light water reactor has been designed. A noveltype of fuel assembly with dual rows of fuel rods between water rods is chosen andoptimized to get more uniform assembly power distributions. Stainless steel is usedfor fuel rod cladding and structural material. Honeycomb structure filled with thermalisolation is introduced to reduce the usage of stainless steel and to keep moderatortemperature below the pseudo critical temperature. Water flow scheme with ascendingcoolant flow in inner regions is carried out to achieve high outlet temperature. In orderto enhance coolant outlet temperature, the radial power distributions needs to be asflat as possible through operation cycle. Fuel loading pattern and control rod patternare optimized to flatten power distribution at inner regions. Axial fuel enrichment isdivided into three parts to control axial power peak, which affects maximum claddingsurface temperature.11:10 AMConceptual Configurations of an Accelerator-Driven SubcriticalSystem Utilizing Minor ActinidesYan Cao and Yousry GoharNuclear Engineering Division, Argonne National Laboratory, Argonne, ILThis paper purposes an Accelerator-Driven Subcritical (ADS) system which utilizesthe Minor Actinides (MAs) from the US spent nuclear fuel inventory. A mobile fuelconcept with micro- particles suspended in the liquid metal is adopted in the purposedsystem to avoid difficulties of developing and testing new MAs solid fuel forms. ThreeADS configurations were developed and analyzed using the Monte Carlo fuel burnupmethodology. The analyses demonstrated the capabilities of the proposed system toutilize the MAs and to dispose of the US spent nuclear fuels.11:35 AMEvaluation of Core Physics Analysis Methods for Conversionof the INL Advanced Test Reactor To Low Enrichment FuelMark D. DeHart and Gray S. ChangIdaho National Laboratory, Idaho Falls, IDComputational neutronics studies to support the possible conversion of the ATR toLEU are underway. Simultaneously, INL is engaged in a physics methods upgradeproject to put into place modern computational neutronics tools for future support ofATR fuel cycle and experiment analysis. A number of experimental measurementshave been performed in the ATRC in support of the methods upgrade project, andare being used to validate the new core physics methods. The current computationalneutronics work is focused on performance of scoping calculations for the ATR coreloaded with a candidate LEU fuel design. This will serve as independent confirmationof analyses that have been performed previously, and will evaluate some of the newcomputational methods for analysis of a candidate LEU fuel for ATR.34

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 10:20 AM - 301 E8C - Advanced Modeling & Simulation in Reactor PhysicsSession Chair: Glenn Sjoden (GT) ; Maria Avramova (PSU)10:20 AMDevelopment of Reactivity Feedback Effect MeasurementTechniques under Sub-critical Condition in Fast ReactorsAkihiro Kitano, Hiroshi Nishi, Takayuki Suzuki (1), Shigeaki Okajima (2),Shigeru Kanemoto (3)1) Japan Atomic Energy Agency, Tsuruga-shi, Fukui-ken, Japan. 2) Japan Atomic Energy Agency, Nakagun,Ibaraki-ken, Japan. 3) The University of Aizu, Aizu-Wakamatsu-shi, Fukushima-ken, JapanThe first-of-a-kind reactor has been licensed by a safety examination of the plant designbased on the measured data in precedent mock-up experiments. The validity ofthe safety design can be confirmed without a mock-up experiment, if the reactor feedbackcharacteristics can be measured before operation, with the constructed reactoritself. The “Synthesis Method”, a systematic and sophisticated method of sub-criticalitymeasurement, is proposed in this work to ensure the safety margin before operation.The “Synthesis Method” is based on the modified source multiplication method (MSM)combined with the noise analysis method to measure the reference sub-criticality levelfor MSM. A numerical simulation for the control-rod reactivity worth and the isothermalfeed-back reactivity was conducted for typical fast reactors of 100MWe-size, 300MWesize,750MWe-size, and 1500MWe-size to investigate the applicability of SynthesisMethod. The number of neutron detectors and their positions necessary for the measurementwere investigated for both methods of MSM and the noise analysis by aseries of parametric survey calculations. As a result, it was suggested that a neutrondetector located above the core center and three or more neutron detectors locatedabove the radial blanket region enable the measurement of sub-criticality within 10%uncertainty from -$0.5 to -$2 and within 15% uncertainty for the deeper sub-criticality.10:45 AMTrace/Parcs Analysis of the OECD/NEA OSKARSHAMN-2BWR Stability BenchmarkTomasz Kozlowski (1), Thomas Downar, Yunlin Xu, Aaron Wysocki (2),Kostadin Ivanov, Jeffrey Magedanz, Matthew Hardgrove (3), Jose March-Leuba (4), Nathanael Hudson, Diana Woodyatt (5)1) University of Illinois at Urbana-Champaign, U.S.A.. 2) University of Michigan, Ann Arbor, Michigan,U.S.A. 3) Pennsylvania State University, University Park, Pennsylvania, U.S.A. 4) Oak Ridge NationalLaboratory, Oak Ridge, Tennessee, U.S.A. 5) Nuclear Regulatory Commission, Rockville, Md, U.S.A.On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event whichculminated in diverging power oscillations with a decay ratio of about 1.4. The eventwas successfully modeled by the TRACE/PARCS coupled code system, and furtheranalysis of the event is described in this paper. The results show very good agreementwith the plant data, capturing the entire behavior of the transient including theonset of instability, growth of the oscillations (decay ratio) and oscillation frequency.This provides confidence in the prediction of other parameters which are not availablefrom the plant records. The event provides coupled code validation for a challengingBWR stability event, which involves the accurate simulation of neutron kinetics (NK),thermalhydraulics (TH), and TH/NK coupling. The success of this work has demonstratedthe ability of the 3-D coupled systems code TRACE/PARCS to capture thecomplex behavior of BWR stability events. The problem was released as an internationalOECD/NEA benchmark, and it is the first benchmark based on measured plantdata for a stability event with a DR greater than one. Interested participants are invitedto contact authors for more information.11:10 AMStatus of Verification & Validation of Areva’s ARCADIA® CodeSystem for PWR ApplicationsDieter Porsch, Mario Leberig , Sebastian Kuch (1), Philippe Magat (2),Kevin Segard (3)1) AREVA, AREVA NP GmbH. 2) AREVA, AREVA NP SAS, Paris. 3) AREVA, AREVA NP INC, LynchburgIn March 2010 the submittal of Topical Reports for ARCADIA® and COBRA-FLX, thethermal-hydraulic module of ARCADIA®, to the U.S. Nuclear Regulatory Commission(NRC) concluded a major step in the development of AREVA’s new code system forcore design and safety analyses. This submittal was dedicated to the application of thecode system to uranium fuel in pressurized water reactors. The submitted informationcomprised results for plants operated in the US, France and Germany and provideduncertainties for in-core measuring systems with traveling in-core detectors and forthe aeroball system of the EPR. A reduction of the uncertainties in the prediction ofFΔH and FQ of > 1 % (absolute) was derived compared to the current code systems.This paper extents the verification and validation base for uranium based fuel anddemonstrates the basic capabilities of ARCADIA® of describing MOX. The achievedstatus of verification and validation is described in detail. All applications followed thesame standard without any specific calibration. The paper gives also insight in the newcapability of 3D full core steady-state and transient pin-by-pin/sub-channel-by-subchannelcalculations and the opportunities offered by this feature. The gain of marginswith increasing detail of the representation is outlined. Currently, the strategies forworldwide implementation of ARCADIA® are developed.11:35 AMFeasibility of Wavelet Expansion Methods to Treat the EnergyVariableW.F.G. van RooijenResearch Institute of Nuclear Engineering, University of Fukui, Tsuruga-shi, Kanawa-cho, JapanThis paper discusses the use of the Discrete Wavelet Transform (DWT) to implementa functional expansion of the energy variable in neutron transport. The motivation ofthe work is to investigate the possibility of adapting the expansion level of the neutronflux in a material region to the complexity of the cross section in that region. If suchan adaptive treatment is possible, “simple” material regions (e.g., moderator regions)require little effort, while a detailed treatment is used for “complex” regions (e.g., fuelregions). Our investigations show that in fact adaptivity cannot be achieved. The mostfundamental reason is that in a multi-region system, the energy dependence of thecross section in a material region does not imply that the neutron flux in that regionhas a similar energy dependence. If it is chosen to sacrifice adaptivity, then the DWTmethod can be very accurate, but the complexity of such a method is higher than thatof an equivalent hyper-fine group calculation. The conclusion is thus that, unfortunately,the DWT approach is not very practical.35

Dr. Jerry HopwoodVice President, Marketing and Product DevelopmentCandu Energy, Inc.Event starts at 12:00 PM“Canadian Response to Fukushima”Introduction by Ron EllisPHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 -12:00 PM - Ballroom A-DLuncheonMr. Hopwood was awarded his Bachelor’s and Masters degrees in Applied Physics from Oxford Universityand then joined the nuclear indusrty in the UK. Recruited by AECL in 1975, Mr. Hopwood’s first specialtywas in Nuclear Safety and Licensing for the Ontario Hydro fleet of units and for the first CANDU 6 designs.From 1983-85, he served as AECL’s on-site team-leader during the initial operation of the CANDU 6 unit at Point Lepreau. In 1992, Mr.Hopwood was appointed Technical Director, South Korea. Returning from Korea in 1995, he was a member of the negotiating team forAECL’s successful bid for the Qinshan CANDU 6 project.Later, Mr. Hopwood served as the Director of Advanced Reactor TechnologyDevelopment, leading the definition of AECL’s Advanced CANDU and EC6 Reactor designs. He was appointed Vice President, ProductDevelopment at AECL in 2005. On the restructuring of AECL’s commercial business to Candu Energy Inc in 2011, Jerry was appointedthe Vice President Marketing and Product Development for Candu Energy Inc.36

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 1:30 PM - 301 A12A - Sensitivity & Uncertainty AnalysisSession Chair: Hany Abdel-Khalik (NCSU); Tatiana Ivanova (IRSN)1:30 PMRe-estimation of Nuclear Data and JEFF3.1.1 UncertaintyCalculationsA. Santamarina, D. Bernard, N. Dos Santos, O. Leray, C. Vaglio and L.LealCommissariat à l’Energie Atomique et aux Energies Alternatives, CEA, DEN, DER, SPRC, Saint-Paul-Lez-Durance, FranceThis paper describes the method to define relevant targeted integral measurementsthat allow the improvement of nuclear data evaluations and the determination of correspondingreliable covariances. 235U and 56Fe examples are pointed out for theimprovement of JEFF3 data. Utilizations of these covariances are shown for Sensitivityand Representativity studies, Uncertainty calculations, and Transposition of experimentalresults to industrial applications.1:55 PMBenchmark On Sensitivity Calculation (Phase III)Tatiana Ivanova, Cédric Laville (1), James Dyrda (2), Dennis Mennerdahl(3), Yury Golovko, Kirill Raskach, Anatoly Tsiboulia (4), Gil Soo Lee, Sweng-Woong Woo (5), Adrien Bidaud, Pouya Sabouri (6), Amrit Patel (7), KeithBledsoe, Bradley Rearden (8), Jim Gulliford, Franco Michel-Sendis (9)1) Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Fontenay aux Roses, France. 2) AtomicWeapons Establishment (AWE), Reading, UK. 3) E Mennerdahl Systems (EMS), Täby, Sweden. 4)Institute for Physics and Power Engineering (IPPE), Obninsk, Russia. 5) Korea Institute of NuclearSafety (KINS), Daejeon, Republic of Korea. 6) Labratoire de Physique Subatomique et de Cosmologie(LPSC),Grenoble, France. 7) U.S. Nuclear Regulatory Commission (NRC), Washington, DC. 8) OakRidge National Laboratory (ORNL), Oak Ridge, TN, USA. 9) OECD/NEA, Issy-les-Moulineaux, FranceThe sensitivities of the keff eigenvalue to neutron cross sections have become commonlyused in similarity studies and as part of the validation algorithm for criticalitysafety assessments. To test calculations of the sensitivity coefficients, a benchmarkstudy (Phase III) has been established by the OECD-NEA/WPNCS/EG UACSA (ExpertGroup on Uncertainty Analysis for Criticality Safety Assessment). This paper presentssome sensitivity results generated by the benchmark participants using variouscomputational tools based upon different computational methods: SCALE/TSUNAMI-3D and -1D, MONK, APOLLO2-MORET 5, DRAGON-SUSD3D and MMKKENO. Thestudy demonstrates the performance of the tools. It also illustrates how model simplificationsimpact the sensitivity results and demonstrates the importance of ‘implicit’(self-shielding) sensitivities. This work has been a useful step towards verification ofthe existing and developed sensitivity analysis methods.2:20 PMA Generalized Adjoint Approach for Quantifying Reflector AssemblyDiscontinuity Factor UncertaintiesArtem Yankov and Benjamin Collins (1), Matthew A. Jessee (2), ThomasDownar (1)1) University of Michigan, Ann Arbor, MI. 2) Oak Ridge National Laboratory, Oak Ridge, TNSensitivity-based uncertainty analysis of assembly discontinuity factors (ADFs) canbe readily performed using adjoint methods for infinite lattice models. However, thereis currently no adjoint-based methodology to obtain uncertainties for ADFs along aninterface between a fuel and reflector region. To accommodate leakage effects in areflector region, a 1D approximation is usually made in order to obtain the homogeneousinterface flux required to calculate the ADF. Within this 1D framework an adjointbasedmethod is proposed that is capable of efficiently calculating ADF uncertainties.In the proposed method the sandwich rule is utilized to relate the covariance of theinput parameters of 1D diffusion theory in the reflector region to the covariance of theinterface ADFs. The input parameters’ covariance matrix can be readily obtained usingsampling-based codes such as XSUSA or adjoint-based codes such as TSUNAMI.The sensitivity matrix is constructed using a fixed-source adjoint approach for inputscharacterizing the reflector region. An analytic approach is then used to determinethe sensitivity of the ADFs to fuel parameters using the neutron balance equation. Astochastic approach is used to validate the proposed adjoint-based method.2:45 PMComparison of XSUSA and “TWO-STEP” Approaches for Full-Core Uncertainty QuantificationArtem Yankov (1), Markus Klein (2), Matthew A. Jessee (3), Winfried Zwermann, KirilVelkov, Andreas Pautz (2), Benjamin Collins, Thomas Downar (1)1) University of Michigan, Ann Arbor, MI. 2) Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH,München, Germany. 3) Oak Ridge National LaboratoryWhile there are multiple sources of error that are introduced into the standard computationalregime for simulating reactor cores, rigorous uncertainty analysis methodsare available primarily for quantifying the effects of cross section uncertainties. Twomethods for propagating cross section uncertainties through core simulators are theXSUSA statistical approach and the “Two-Step” method. The XSUSA approach, whichis based on the SUSA code package, is fundamentally a stochastic sampling method.Alternatively, the Two-Step method utilizes generalized perturbation theory in thefirst step and stochastic sampling in the second step. The consistency of these twomethods in quantifying uncertainties in the multiplication factor and in the core powerdistribution will be examined in the framework of phase I-3 of the UAM Benchmark.Using the TMI core as a base model for analysis, the XSUSA and Two-Step methodsare applied with certain limitations and the results are compared to those producedby other stochastic sampling-based codes. Based on the uncertainty analysis results,conclusions are made for which method is currently a more viable option for computinguncertainties in burnup and transient calculations.3:10 PMDevelopment of a Statistical Sampling Method for Uncertaintyanalysis with ScaleM. Williams, D. Wiarda, H. Smith, M. A. Jessee, B. T. Rearden (1), W. Zwermann,M. Klein, A. Pautz, B. Krzykacz-Hausmann, L. Gallner (2)1) Oak Ridge National Laboratory, Oak Ridge, TN, USA. 2) Gesellschaft fuer Anlagen- und Reaktorsicherheit(GRS), Garching, GermanyA new statistical sampling sequence called Sampler has been developed for theSCALE code system. Random values for the input multigroup cross sections are determinedby using the XSUSA program to sample uncertainty data provided in theSCALE covariance library. Using these samples, Sampler computes perturbed selfshieldedcross sections and propagates the perturbed nuclear data through any specifiedSCALE analysis sequence, including those for criticality safety, lattice physicswith depletion, and shielding calculations. Statistical analysis of the output distributionsprovides uncertainties and correlations in the desired responses.3:35 PMAleatoric and Epistemic Uncertainties in Sampling BasedNuclear Data Uncertainty and Sensitivity AnalysesW. Zwermann, B. Krzykacz-Hausmann, L. Gallner, M. Klein, A. Pautz, K.VelkovGesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH Garching, GermanySampling based uncertainty and sensitivity analyses due to epistemic input uncertainties,i.e. to an incomplete knowledge of uncertain input parameters, can be performedwith arbitrary application programs to solve the physical problem under consideration.For the description of steady-state particle transport, direct simulations of the microscopicprocesses with Monte Carlo codes are of- ten used. This introduces an additionalsource of uncertainty, the aleatoric sampling uncertainty, which is due to therandomness of the simulation process performed by sampling, and which adds to thetotal combined output sampling uncertainty. So far, this aleatoric part of uncertainty ismini- mized by running a sufficiently large number of Monte Carlo histories for eachsample calculation, thus making its impact negligible as compared to the impact fromsampling the epistemic uncer- tainties. Obviously, this process may cause high computationalcosts. The present paper shows that in many applications reliable epistemicuncertainty results can also be obtained with substantially lower computational effortby performing and analyzing two appropriately generated series of samples with muchsmaller number of Monte Carlo histories each. The method is applied along with thenuclear data uncertainty and sensitivity code package XSUSA in combination with theMonte Carlo transport code KENO-Va to various critical assemblies and a full scalereactor calcu- lation.4:00 PMInterpolations of Nuclide-Specific Scattering Kernels Generatedwith SerpentAnthony Scopatz and Erich SchneiderThe University of Texas at Austin, Austin, TXThe neutron group-to-group scattering cross section is an essential input parameterfor any multi-energy group physics model. However, if the analyst prefers to use MonteCarlo transport to generate group constants this data is difficult to obtain for a singlespecies of a material. Here, the Monte Carlo code Serpent was modified to returnthe group transfer probabilities on a per-nuclide basis. This ability is demonstrated inconjunction with an essential physics reactor model where cross section perturbationsare used to dynamically generate reactor state dependent group constants via interpolationfrom pre-computed libraries. The modified version of Serpent was thereforeverified with three interpolation cases designed to test the resilience of the interpolationscheme to changes in intragroup fluxes. For most species, interpolation resultedin errors of less than 5% of transport-computed values. For important scatterers, suchas 1H, errors less than 2% were observed. For nuclides with high errors (> 10%), thescattering channel typically only had a small probability of occurring.37

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 1:30 PM - 301 ASS1A - Special Session in honor of Nils Göran SjöstrandSession Chair: Imre Pazsit (Chalmers); Dimitrios Cokinos (BNL)Special Session in honor of Nils Göran SjöstrandThe 2011 winner of the ANS Reactor Physics Division’s Eugene P. Wigner Reactor Physicist AwardThe ANS Reactor Physics Division’s (RPD) Honors and Awards Committee selected Nils.G. Sjöstrand, Professor Emeritus at Chalmers University of Technology,as the 2011 winner of the Eugene P. Wigner Reactor Physicist Award. This Special Session is dedicated to Dr. Sjöstrand.Since Dr. Sjöstrand could not travel to the U.S. to receive his Wigner Award plaque, a special ceremony was organized at the Chalmers University ofTechnology in Göteborg, Sweden, on November 19, 2011. In lieu of the Wigner Keynote Address, the PHYSOR 2012 technical program chair and organizationalcommittee and RPD decided to pay tribute to Professor Sjöstrand by dedicating this Special Session to his name. A discussion of the highlights ofSjöstrand’s major accomplishments will be presented by Imre Pazsit, followed by a brief report by Dimitrios Cokinos on the award ceremony that took placein Göteborg. The Special Session will continue with the presentation of a set of papers, each subject area related to some of Dr. Sjöstrand’s work.1:30 PMNils Göran Sjöstrand’s contributions to nuclear engineering - portrait of a scientist and a private personby Imre PázsitThe start of Nils Göran Sjöstrand’s professional career coincides with the early development of reactor physics and nuclear power research and educationin Sweden. The first research reactor, R1, was started up in central Stockholm at the campus of the Royal Institute of Technology where Nils Göran did hisgraduate studies, while being employed the newly founded AB Atomenergi, the governmental research institute for the development of nuclear power inSweden. Sigvard Eklund, later the General Director of the IAEA, was the head of the research group at Atomenergi, and the advisor of N. G. Sjöstrand.Part of his PhD work, during which he collaborated extensively with Guy von Dardel, was the elaboration of the area ratio method, for which he is mostknown internationally.This talk will give an account of Nils Göran Sjöstrand’s work during the pioneering times of nuclear energy in Sweden. His contributions, which constitutedthe motivation for the Wigner Award in the nominating letter, will be listed. These will be mixed with numerous stories and anecdotes from those ‘early days’,told by Nils Göran himself and noted down by the present speaker (such as when he performed the secret mission of escorting 3 tons of heavy water fromNorway for the building of the R1 reactor in Stockholm).Imre will mix this talk with further impressions and reminiscences, based on his personal experience as the successor of Nils Göran as professor and Chairof the department at Chalmers, having spent over 20 years’ time together at the department while he has been emeritus.2:20 PMSpatial Correction Factors for YALINA Booster Facility Loadedwith Medium and Low Enriched FuelsAlberto Talamo, Yousry Gohar (1), V. Bournos, Y. Fokov, H. Kiyavitskaya,C. Routkovskaya (2)1) Argonne National Laboratory, Argonne, IL, U.S.A. 2) Joint Institute for Power and Nuclear Research-Sosny, Minsk, BelarusBell and Glasstone spatial correction factor is used in analyses of subcritical assembliesto correct the experimental reactivity as function of the detector position. Besidesthe detector position, several other parameters affect the correction factor: the energyweighting function of the detector, the detector size, the energy-angle distribution ofsource neutrons, and the reactivity of the subcritical assembly. This work focuses onthe dependency of the correction factor on the detector material and it investigatesthe YALINA Booster subcritical assembly loaded with medium (36%) and low (10%)enriched fuels.2:45 PMNeural Network and Area Method Interpretation of Pulsed ExperimentsS. Dulla, P. Picca, P. Ravetto (1), S. Canepa (2)1) Politecnico di Torino, Dipartimento di Energetica, Torino, Italy. 2) Lab of Reactor Physics & SystemsBehaviour (LRS), Paul Scherrer Institute, Villigen, Switzerland3:10 PMNeutron Noise Measurements at the Delphi Subcritical AssemblyMáté Szieberth and Gergely Klujber (1), Jan Leen Kloosterman and Dickde Haas (2)Institute of Nuclear Techniques, Budapest University of Technology and Economics (BME), Műegyetem,Hungary. 2) Section Physics of Nuclear Reactors, Delft University of Technology (TUD), DelftThe paper presents the results and evaluations of a comprehensive set of neutronnoise measurements on the Delphi subcritical assembly of the Delft University ofTechnology. The measurements investigated the effect of different source distributions(inherent spontaneous fission and 252Cf) and the position of the detectors applied(both radially and vertically). The evaluation of the measured data has been performedby the variance-to-mean ratio (VTMR, Feynman-), the autocorrelation (ACF, Rossi-)and the cross-correlation (CCF) methods. The values obtained for the prompt decayconstant show a strong bias, which depends both on the detector position and onthe source distribution. This is due to the presence of higher modes in the system.It has been observed that the value fitted is higher when the detector is close to theboundary of the core or to the 252Cf point-source. The higher alpha-modes have alsobeen observed by fitting functions describing two alpha-modes. The successful set ofmeasurement also provides a good basis for further theoretical investigations includingthe Monte Carlo simulation of the noise measurements and the calculation of thealpha-modes in the Delphi subcritical assembly.The determination of the subcriticality level is an important issue in accelerator-drivensystem technology. The area method, originally introduced by N. G. Sjöstrand, is aclassical technique to interpret flux measurement for pulsed experiments in order toreconstruct the reactivity value. In recent times other methods have also been developed,to account for spatial and spectral effects, which were not included in the areamethod, since it is based on the point kinetic model. The artificial neural network approachcan be an efficient technique to infer reactivities from pulsed experiments. Inthe present work, some comparisons between the two methods are carried out anddiscussed.38

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 1:30 PM - 301 ASS1A - Special Session in honor of Nils Göran SjöstrandContinued3:35 PMDetector Positioning for the Initial Subcriticality Level Determinationin Accelerator-Driven SystemsW. Uyttenhove, G. van den Eynde, P. Baeten, A. Kochetkov, G. Vittiglioand J.Wagemans (1), D. Lathouwers, J.-L. Kloosterman, T.J.H.H. van derHagen and F.Wols (2), A. Billebaud, S. Chabod and H.-E. Thybault (3), J.-L.Lecouey, G. Ban, F.-R. Lecolley, N. Marie and J.-C. Steckmeyer (4), P. Dessagneand M. Kerveno (5), F. Mellier (6)1) SCK•CEN, Belgian Nuclear Research Centre, Mol, Belgium. 2) Delft University of Technology, Delft,The Netherlands. 3) LPSC-CNRS-IN2P3/UJF/INPG, Grenoble cedex, France. 4) LPC Caen, ENSICAEN/Unicersit de Caen/CNRS-IN2P3, Caen, France. 5) IPHC-DRS/UdS/CNRS-IN2P3, Strasbourg, France. 6)CEA/DEN/DER/SPEX/LPE Cadarache, Saint-Paul-les-Durance, FranceWithin the GUINEVERE project (Generation of Uninterrupted Intense NEutrons at thelead VEnus REactor) carried out at SCK•CEN in Mol, the continuous deuteron acceleratorGENEPI-3C was coupled to the VENUS-F fast simulated lead-cooled reactor.Today the FREYA project (Fast Reactor Experiments for hYbrid Applications) is ongoingto study the neutronic behavior of this Accelerator Driven System (ADS) duringdifferent phases of operation. In particular the set-up of a monitoring system for thesubcriticality of an ADS is envisaged to guarantee safe operation of the installation.W. Uyttenhove et al. The methodology for subcriticality monitoring in ADS takes intoaccount the determination of the initial subcriticality level, the monitoring of reactivityvariations, and interim cross-checking. At start-up, the Pulsed Neutron Source (PNS)technique is envisaged to determine the initial subcriticality level. Thanks to its referencecritical state, the PNS technique can be validated on the VENUS-F core. Adetector positioning methodology for the PNS technique is set up in this paper forthe subcritical VENUS-F core, based on the reduction of higher harmonics in a staticevaluation of the Sjöstrand area method. A first case study is provided on the VENUS-F core. This method can be generalised in order to create general rules for detectorpositions and types for full-scale ADS.4:00 PMGUINEVERE Experiment: Kinetic Analysis of Some ReactivityMeasurement Methods by Deterministic and Monte CarloCodes.G. Bianchini, N. Burgio, M. Carta (1), V. Peluso (2), V. Fabrizio, L. Ricci (3)1) ENEA C.R. CASACCIA, Roma, ITALY. 2) ENEA C.R. BOLOGNA, Bologna, ITALY. 3) University ofRome “La Sapienza”, c/o ENEA C.R. CASACCIA, Roma, ITALYThe GUINEVERE experiment (Generation of Uninterrupted Intense NEutrons at thelead VEnus REactor) is an experimental program in support of the ADS technologypresently carried out at SCK•CEN in Mol (Belgium). In the experiment a modified layoutof the original thermal VENUS critical facility is coupled to an accelerator, built bythe French body CNRS in Grenoble, working in both continuous and pulsed modeand delivering 14 MeV neutrons by bombardment of deuterons on a tritium-target.The modified lay-out of the facility consists of a fast subcritical core made of 30%U235 enriched metallic Uranium in a lead matrix. Several off-line and on-line reactivitymeasurement techniques will be investigated during the experimental campaign.This report is focused on the simulation by deterministic (ERANOS French code) andMonte Carlo (MCNPX US code) calculations of three reactivity measurement techniques,Slope (α-fitting), Area-ratio and Source-jerk, applied to a GUINEVERE subcriticalconfiguration (namely SC1). The inferred reactivity, in dollar units, by the Area-ratiomethod shows an overall agreement between the two deterministic and Monte Carlocomputational approaches, whereas the MCNPX Source-jerk results are affected bylarge uncertainties and allow only partial conclusions about the comparison. Finally,no particular spatial dependence of the results is observed in the case of the GUINE-VERE SC1 subcritical configuration.39

40PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 1:30 PM - 301 C3C - Monte Carlo Methods & DevelopmentsSession Chair: Tom Sutton (KAPL); Dan Gill (Bettis)1:30 PMMonte Carlo Interpretation Of The AMMON/REF ExperimentIn EOLE for the JHR Reactor CalculationsC. Vaglio-Gaudard, O. Leray, A.C. Colombier, C. D’Aletto, L. Gaubert, O.Gueton, J.P. Hudelot, M. Valentini (1), P. Siréta (3), J. Di Salvo, A. Gruel,J.C. Klein, A. Roche, D. Beretz, J.M. Girard (2)1) CEA, DEN, DER/SPRC, Cadarache, Saint Paul les Durance, France. 2) CEA, DEN, DER/SPEx,Cadarache, Saint Paul les Durance, France. 3) CEA, DEN, DER/SRJH, Cadarache, Saint Paul les Durance,FranceA new experiment, named AMMON and dedicated to the analysis of the JHR neutronphysics and to the qualification of the associated neutron calculation tools, is currentlyin progress in the EOLE zero-power experimental reactor at CEA Cadarache. Thefirst core configuration, so called “reference configuration” (AMMON/REF) has alreadybeen loaded: it consists of an experimental zone of 7 JHR assemblies with U3Si227% 235U enriched fuel curved plates surrounded by a driver zone with 623 standardPWR UOx fuel pins. The analysis of the AMMON/REF measurements is presented inthis paper. It is based on calculations performed with the 3-dimensionnal referenceMonte Carlo TRIPOLI4.7 code and the JEFF3.1.1 European library: it highlights agood agreement between calculation and experiment concerning reactivity and powerdistribution in the experimental zone. Reactivity prediction is very satisfactory, despitethe presence of a large aluminum quantity in the core: C-E=+ 376 pcm ± 290. Theradial and axial plate power profiles are also in good agreement. Assembly powerdistribution is predicted within the experimental uncertainties (1s).1:55 PMBurnup Calculationmethodology in the Serpent 2 Monte CarloCodeJaakko Leppänen (1), Aarno Isotalo (2)1) VTT Technical Research Centre of Finland, FI-02044 VTT, Finland. 2) Aalto University, Department ofApplied Physics, FI-00076 AALTO, FinlandThis paper presents two topics related to the burnup calculation capabilities in the Serpent2 Monte Carlo code: advanced time-integration methods and improved memorymanagement, accomplished by the use of different optimization modes. The developmentof the introduced methods is an important part of re-writing the Serpent sourcecode, carried out for the purpose of extending the burnup calculation capabilities from2D assembly-level calculations to large 3D reactor-scale problems. The progress isdemonstrated by repeating a PWR test case, originally carried out in 2009 for the validationof the newly-implemented burnup calculation routines in Serpent 1.2:20 PMIs Monte Carlo Embarrassingly Parallel?J. Eduard Hoogenboom (1,2)1) Delft University of Technology, Delft, The Netherlands. 2) Delft Nuclear Consultancy, Capelle aan denIJssel, The NetherlandsMonte Carlo is often stated as being embarrassingly parallel. However, running aMonte Carlo calculation, especially a reactor criticality calculation, in parallel usingtens of processors shows a serious limitation in speedup and the execution time mayeven increase beyond a certain number of processors. In this paper the main causesof the loss of efficiency when using many processors are analyzed using a simpleMonte Carlo program for criticality. The basic mechanism for parallel execution is MPI.One of the bottlenecks turn out to be the rendez-vous points in the parallel calculationused for synchronization and exchange of data between processors. This happens atleast at the end of each cycle for fission source generation in order to collect the fullfission source distribution for the next cycle and to estimate the effective multiplicationfactor, which is not only part of the requested results, but also input to the next cycle forpopulation control. Basic improvements to overcome this limitation are suggested andtested. Also other time losses in the parallel calculation are identified. Moreover, thethreading mechanism, which allows the parallel execution of tasks based on sharedmemory using OpenMP, is analyzed in detail. Recommendations are given to get themaximum efficiency out of a parallel Monte Carlo calculation.2:45 PMAn Efficient Implementation of the Chebyshev Rational ApproximationMethod (CRAM) for Solving the Burnup EquationsMaria Pusa and Jaakko LeppänenVTT Technical Research Centre of Finland, FI-02044 VTT, FinlandThe Chebyshev Rational Approximation Method (CRAM) has been recently introducedby the authors for solving the burnup equations with excellent results. This method hasbeen shown to be capable of simultaneously solving an entire burnup system withthousands of nuclides both accurately and efficiently. The method was prompted by ananalysis of the spectral properties of burnup matrices and it can be characterized asthe best rational approximation on the negative real axis. The coefficients of the rationalapproximation are fixed and have been reported for various approximation orders.In addition to these coefficients, implementing the method only requires a linear solver.This paper describes an efficient method for solving the linear systems associated withthe CRAM approximation. The introduced direct method is based on sparse Gaussianelimination where the sparsity pattern of the resulting upper triangular matrix is determinedbefore the numerical elimination phase. The stability of the proposed Gaussianelimination method is discussed based on considering the numerical properties of burnupmatrices. Suitable algorithms are presented for computing the symbolic factorizationand numerical elimination in order to facilitate the implementation of CRAM andits adoption into routine use. The accuracy and efficiency of the described techniqueare demonstrated by computing the CRAM approximations for a large test case withover 1600 nuclides.3:10 PMNumerical Study of Error Propagation in Monte Carlo DepletionSimulationsTimothy Wyant and Bojan PetrovicNuclear and Radiological Engineering, Georgia Institute of Technology, Atlanta, GAImproving computer technology and the desire to more accurately model the heterogeneityof the nuclear reactor environment have made the use of Monte Carlo depletioncodes more attractive in recent years, and feasible (if not practical) even for 3-D depletionsimulation. However, in this case statistical uncertainty is combined with errorpropagating through the calculation from previous steps. In an effort to understandthis error propagation, a numerical study was undertaken to model and track individualfuel pins in four 17x17 PWR fuel assemblies. By changing the code’s initial randomnumber seed, the data produced by a series of 19 replica runs was used to investigatethe true and apparent variance in k-eff, pin powers, and number densities of severalisotopes. While this study does not intend to develop a predictive model for errorpropagation, it is hoped that its results can help to identify some common regularitiesin the behavior of uncertainty in several key parameters.3:35 PMPreliminary TRIGA Fuel Burn-Up Evaluation by Means ofMonte Carlo Code and Computation Based on Total EnergyReleased During Reactor OperationA. Borio di Tigliole, J. Bruni, F. Panza (1), D. Alloni, M. Cagnazzo, G.Magrotti, S. Manera, M. Prata,A. Salvini (2), D. Chiesa, M. Clemenza, L.Pattavina, E. Previtali, M. Sisti (3), A. Cammi (4)1) Department of Nuclear and Theoretical Physics - University of Pavia and Italian National Institute ofNuclear Physics (INFN) - Section of Pavia, Pavia, Italy. 2) Applied Nuclear Energy Laboratory (LENA)- University of Pavia, Pavia, Italy and Italian National Institute of Nuclear Physics (INFN) - Section ofPavia. 3) Physics Department “G. Occhialini” - University of Milano Bicocca and Italian National Instituteof Nuclear Physics (INFN) - Section of Milano Bicocca, Milano, Italy. 4) Department of Energy “EnricoFermi Centre for Nuclear Studies (CeSNEF)”, Polytechnic University of Milan, Milano (Italy) and ItalianNational Institute of NuclearAim of this work was to perform a rough preliminary evaluation of the burn-up of thefuel of TRIGA Mark II research reactor of the Applied Nuclear Energy Laboratory(LENA) of the University of Pavia. In order to achieve this goal a computation of theneutron flux density in each fuel element was performed by means of Monte Carlocode MCNP (Version 4C). The results of the simulations were used to calculate the effectivecross sections (fission and capture) inside fuel and, at the end, to evaluate theburn-up and the uranium consumption in each fuel element. The evaluation, showeda fair agreement with the computation for fuel burn-up based on the total energy releasedduring reactor operation.4:00 PMThe 3-D Monte Carlo Neutron and Photon Transport CodeMCMG and Its AlgorithmsDENG Li, HU Zehua, LI Gang, LI Shu, and LIU ZhengzhouThe Institute of Applied Physics and Computational Mathematics, Beijing, P.R. ChinaThe 3-D Monte Carlo neutron and photon transport parallel code MCMG is developed.A new collision mechanism based on material but not nuclide is added in the code.Geometry cells and surfaces can be dynamically extended. Combination of multigroupand continuous cross-section transport is developed. The multigroup scattering is expansibleto P5 and upper scattering is considered. Various multigroup libraries can beeasily equipped in the code. The same results with the experiments and the MCNPcode are obtained for a series of modes. The speedup of MCMG is a factor of 2-4 relativeto the MCNP code in speed.

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 1:30 PM - 301 D16A - Radiation Applications & Nuclear SafeguardsSession Chair: Sara Pozzi (U Mich); Tim Valentine (ORNL)1:30 PMNoninvasive Cross Section Reconstruction with TransportTheory ConstraintsNathaniel Fredette, Jean Ragusa and Wolfgang BangerthTexas A&M University, College Station, TXWe consider the inverse problem of identifying the spatially variable absorption andscattering properties of a medium by measuring the exiting radiation when the body isactively interrogated. We formulate this inverse problem as a PDE-constrained optimizationproblem and solve it iteratively with Newton’s method. The constraint is givenby the radiative transport equation for neutral particles. Two examples are considered.The first is a dual inclusion domain with no scattering. This problem explores the convergencepatterns of the method. The second problem is a central inclusion problemwith scattering. This problem explores the optical thickness limit of the method. Thisoptical thickness was determined to be 2-3 mean free paths.1:55 PMMCNP6 Enhancements of Delayed-Particle ProductionGregg W. McKinneyLos Alamos National Laboratory, Los Alamos, NMOver the last decade, there has been an increased interest in the production ofdelayed-particle signatures from neutron and photon interactions with matter. To addressthis interest, various radiation transport codes have developed a wide range ofdelayed-particle physics packages. With the recent merger of the Monte Carlo transportcodes MCNP5 and MCNPX, MCNP6 inherited the comprehensive model-baseddelayed-particle production capabilities developed in MCNPX over the last few years.An integral part of this capability consists of the depletion code CINDER90 which wasincorporated into MCNPX in 2004. During this last year, significant improvements havebeen made to the MCNP6 physics and algorithms associated with delayed-particleproduction, including the development of a delayed-beta capability, an algorithm enhancementfor the delayed-neutron treatment, and a database enhancement for delayed-gammaemission. The delayed-beta feature represents an important componentin modeling background signals produced by active interrogation sources. Combined,these improvements provide MCNP6 with a flexible state-of-the-art physics packagefor generating high-fidelity signatures from fission and activation. This paper providesdetails of these enhancements and presents results for a variety of fission and activationexamples.the total neutron counts over the entire energy spectrum. Other Blocks detect differentneutron energies. All five neutron detector blocks and the gamma-ray block are assembledin both MCNP and deterministic simulation models, with detector responsescalculated to validate the fully assembled design using a 30-group library. The simulationresults show that the 30-group library, collapsed from an 80-group library using anadjoint-weighting approach with the YGROUP code, significantly reduced the computationalcost while maintaining accuracy.3:10 PMOrganic Scintillation Detector Response Simulation UsingNon-Analog MCNPX-POLIMIShikha Prasad, Shaun D. Clarke, Sara A. Pozzi and Edward W. LarsenUniversity of Michigan, Ann Arbor, MIOrganic liquid scintillation detectors are valuable for the detection of special nuclearmaterial since they are capable of detecting both neutrons and gamma rays. Scintillatorscan also provide energy information which is helpful in identification and characterizationof the source. In order to design scintillation based measurement systemsappropriate simulation tools are needed. MCNPXPoliMi is capable of simulating scintillationdetector response; however, simulations have traditionally been run in analogmode which leads to long computation times. In this paper, nonanalog MCNPX-PoliMimode which uses variance reduction techniques is applied and tested. The non-analogMCNPX-PoliMi simulation test cases use source biasing, geometry splitting and acombination of both variance reduction techniques to efficiently simulate pulse heightdistribution and then time-of-flight for a heavily shielded case with a 252Cf source.An improvement factor (I), is calculated for distributions in each of the three casesabove to analyze the effectiveness of the non-analog MCNPX-PoliMi simulations inreducing computation time. It is found that of the three cases, the last case which usesa combination of source biasing and geometry splitting shows the most improvementin simulation run time for the same desired variance. For pulse height distributionsspeedup ranging from a factor 5 to 25 is observed, while for time-of-flights the speedupfactors range from 3 to 10.3:35 PMInverse Transport Problem Solvers Based on Regularized andCompressive Sensing TechniquesYuxiong Cheng, Liangzhi Cao, Hongchun Wu, Hongbo ZhangSchool of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shaanxi, China2:20 PMSimultaneous 233U and 235U Characterization Through theAssay of Delayed Neutron Temporal BehaviorM.T. Sellers, E.C. Corcoran, D.G. KellyDepartment of Chemistry and Chemical Engineering, Royal Military College of Canada, Ontario, CanadaAqueous solutions containing dissolved uranium-233 and uranium-235 were irradiatedfor 60s in the SLOWPOKE-2 reactor at the Royal Military College of Canada.The temporal behavior of the delayed neutrons produced was recorded by the Facility’sDelayed Neutron Counting (DNC) system. The percentage of uranium-233 as afunction of total fissile mass present in each sample ranged from 0 to 100% and waspredicted by the DNC system with average absolute errors of ± 4%. Future work willupgrade the system electronics and software to reduce both uncertainties in timingsand electrical noise. Mixture analysis will also be expanded to include plutonium-239and fissile materials contained in non-aqueous matrices.According to the direct exposure measurements from flash radiographic image, regularized-basedmethod and compressive sensing (CS)-based method for inverse transportequation are presented. The linear absorption coefficients and interface locationsof objects are reconstructed directly at the same time. With a large number of measurements,least-square method is utilized to complete the reconstruction. Owing tothe ill-posedness of the inverse problems, regularized algorithm is employed. Tikhonovmethod is applied with an appropriate posterior regularization parameter to get ameaningful solution. However, it’s always very costly to obtain enough measurements.With limited measurements, CS sparse reconstruction technique Orthogonal MatchingPursuit (OMP) is applied to obtain the sparse coefficients by solving an optimizationproblem. This paper constructs and takes the forward projection matrix rather thanGauss matrix as measurement matrix. In the CS-based algorithm, Fourier expansionand wavelet expansion are adopted to convert an underdetermined system to a wellposedsystem. Simulations and numerical results of regularized method with appropriateregularization parameter and that of CS-based agree well with the reference value,furthermore, both methods avoid amplifying the noise.2:45 PMOptimization of a Neutron Detector Design Using AjointTransport SimulationCe Yi, Kevin Manalo, Mi Huang, Michael Chin, Christopher Edgar, SpencerApplegate, and Glenn SjodenGeorgia Institute of Technology, Atlanta, GAA synthetic aperture approach has been developed and investigated for Special NuclearMaterials (SNM) detection in vehicles passing a checkpoint at highway speeds.SNM is postulated to be stored in a moving vehicle and detector assemblies areplaced on the road-side or in chambers embedded below the road surface. Neutronand gamma spectral awareness is important for the detector assembly design besideshigh efficiencies, so that different SNMs can be detected and identified with variouspossible shielding settings. The detector assembly design is composed of a CsI gamma-raydetector block and five neutron detector blocks, with peak efficiencies targetingdifferent energy ranges determined by adjoint simulations. In this study, formulationsare derived using adjoint transport simulations to estimate detector efficiencies. Theformulations is applied to investigate several neutron detector designs for Block IV,which has its peak efficiency in the thermal range, and Block V, designed to maximize41

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 1:30 PM - 301 E9A - Research Reactors & Spallation SourcesSession Chair: C. Pyeon (Kyoto U)1:30 PMPerformance and Safety Parameters for the High Flux IsotopeReactorGermina Ilas and Trent Primm IIIOak Ridge National Laboratory, Oak Ridge, TN, USAA Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400and its use in calculating parameters of relevance to the reactor performance andsafety during the reactor cycle are presented in this paper. This depletion model wasdeveloped to serve as a reference for the design of a low-enriched uranium (LEU) fuelfor an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel;both HEU and LEU depletion models use the same methodology and ENDF/B-VIInuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters,which are compared with measurement data from critical experiments performed atHFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported byprevious calculations, provide a basis for verification or updating of the correspondingSAR data.1:55 PMCalculation of Heating Values for the High Flux Isotope ReactorJoshua Peterson and Germina IlasOak Ridge National Laboratory, Oak Ridge, TN, USACalculating the amount of energy released by a fission reaction (fission Q value) andthe heating rate distribution in a nuclear reactor is an important part of the safety analysis.However, these calculations can become very complex. One of the codes that canbe used for this type of analyses is the Monte Carlo transport code MCNP5. Currentlyit is impossible to calculate the Q value and heating rate disposition for delayed betaand delayed gamma particles directly from MCNP5. The purpose of this paper is tooutline a rigorous method for indirectly calculating the Q values and heating rates inthe High Flux Isotope Reactor (HFIR), based on previous similar studies carried outfor very high-temperature reactor configurations. This method has been applied inthis study to calculate heating rates for the beginning of cycle (BOC) and end-of-cycle(EOC) states of HFIR. In addition, the BOC results obtained for HFIR are comparedwith corresponding results for the Advanced Test Reactor. The fission Q value forHFIR was calculated as 200.2 MeV for the BOC and 201.3 MeV for the EOC. It wasalso determined that 95.1% and 95.4% of the heat was deposited within the HFIR fuelplates for the BOC and EOC models, respectively. This methodology can also be usedfor heating rate calculations for HFIR experiments.2:20 PMExperimental and Computational Study of the Flux Spectrumin Materials Irradiation Facilities of the High Flux Isotope ReactorThomas Daly (1), Joel McDuffee (2)1) Department of Nuclear Engineering, University of Tennessee, Knoxville, TN. 2) Oak Ridge NationalLaboratory, Oak Ridge, TNThis report compares the available experimental neutron flux data in the High FluxIsotope Reactor (HFIR) to computational models of the HFIR loosely based on theexperimental loading of Cycle 400. Over the last several decades, many materialsirradiation experiments have included fluence monitors that were subsequently usedto reconstruct a coarse-group energy-dependent flux spectrum. Experimental valuesfor thermal (E < 0.5 eV) and fast (E > 0.1 MeV) neutron flux in the outer ring of the fluxtrap about the midplane are found to be 1:73 0:20 and 1:06 0:04 1015 n cm2sec, respectively. The reactor physics code MCNP is used to calculate neutron flux inthe HFIR at irradiation locations. The computational results are shown to correspondclosely to experimental data for thermal and fast neutron flux with calculated percentdifferences ranging from 0:55{13:20%.2:45 PMThe Effects of Flux Spectrum Perturbation on TransmutationOf Actinides: Optimizing the Production of Transcurium IsotopesSusan Hogle and G. Ivan Maldonado (1), Charles Alexander (2)Department of Nuclear Engineering, University of Tennessee, Knoxville, TN. 2) Oak Ridge National Laboratory,Oak Ridge, TNThe research presented herein involves the optimization of transcurium isotope productionin the High Flux Isotope Reactor at Oak Ridge National Laboratory. Due tothe strong dependence of isotopic cross sections upon incoming neutron energy, theefficiency with which an isotope is transmuted is highly dependent upon the neutronflux energy spectrum and intensities. There are certain energy ranges in which therate of fissions in feedstock materials can be minimized, relative to the rate of nonfissionabsorptions. Therefore, it is proposed that by perturbing the flux spectrum, it ispossible to increase the production yield key isotopes for the heavy element researchprogram, such as 249Bk and 252Cf, which are produced during a transmutation cycle,relative to the consumption of feedstock material. The optimization process is carriedout by developing an iterative objective framework that involves problem definition, fluxspectrum and cross-section analysis, simulated transmutation, and analysis of finalyields and transmutation parameters. It is shown that it is possible to perturb the localflux spectrum in the transcurium target by perturbing the composition of the target. It isfurther shown that these perturbations can alter the target yields in a significant way,increasing the amount of 252Cf produced per feedstock consumption by over 8%.Future work is necessary to develop the optimization framework and to identify thenecessary algorithms to update the problem definition based upon progress towardthe production of the desired transcurium isotope.3:10 PMReanalysis of The Gas-Cooled Fast Reactor Experiments atthe Zero Power Facility Proteus – Spectral IndicesGregory Perret, Rajesh M. Pattupara (1), Gaëtan Girardin, Rakesh Chawla(2)1) Paul Scherrer Institute, Villigen, Switzerland. 2) Ecole Polytechnique Fédérale de Lausanne, Lausanne,SwitzerlandThe gas-cooled fast reactor (GCFR) concept was investigated experimentally in thePROTEUS zero power facility at the Paul Scherrer Institute during the 1970’s. Theexperimental program was aimed at neutronics studies specific to the GCFR and atthe validation of nuclear data in fast spectra. A significant part of the program usedthorium oxide and thorium metal fuel either distributed quasi-homogeneously in thereference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones.Experimental results obtained at the time are still of high relevance in view of the currentconsideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclearsystem, as also of the renewed interest in the thorium cycle. In this context, some ofthe experiments have been modeled with modern Monte Carlo codes to better accountfor the complex PROTEUS whole-reactor geometry and to allow validating recent continuousneutron crosssection libraries. As a first step, the MCNPX model was used totest the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectralindices, notably involving fission and capture of 232Th and 237Np, measured in GFRlikelattices.42

ContinuedPHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 1:30 PM - 301 E9A - Research Reactors & Spallation Sources3:35 PMPower Transient Analyses of Experimental In-Reflector DevicesDuring Safety Shutdown in Jules Horowitz Reactor (JHR)Patrizio Console Camprini and Marco Sumini (1), Carlo Artioli (2), ChristianGonnier, Bernard Pouchin and Serge Bourdon (3)1) University of Bologna, Italy. 2) National Agency for New Technologies, Energy and Sustainable EconomicDevelopment (ENEA), Italy. 3) Atomic Energy Commission (CEA), FranceThe Jules Horowitz Reactor (JHR) is designed to be a 100 MW material testing reactor(MTR) and it is expected to become the reference facility in the framework of Europeannuclear research activity. As the core neutron spectrum is quite fast, several experimentaldevices concerning fuel studies have been conceived to be placed in the reflectorin order to exploit a proper thermal neutron flux irradiation. Since the core power isrelatively high, the neutronic coupling between the reactor core and the reflector deviceshas to be taken into account for different rod insertions. In fact the thermal powerproduced within the fuel samples is considerable. Heat removal during shutdown isa main topic in nuclear safety and it is worth to analyse thermal power transientsin fuel samples as well. Here a thermal hydraulic model for JHR core is proposedaiming at a simple and representative description as far as reactivity feedbacks areconcerned. Then it is coupled with a neutronic pointwise kinetics analysis by means ofthe DULCINEE code to compute core power transient calculations. Moreover, somereflector-core coupling evaluations are performed through Monte Carlo method usingthe TRIPOLI 4.7 code. The JHR equilibrium cycle is considered with respect to fourfuel compositions namely Beginning of Cycle (BOC), Xenon Saturation Point (XSP),Middle of Cycle (MOC) and End of Cycle (EOC). Then thermal power transients in theexperimental reflector devices are evaluated during safety shutdowns and they areverified for all these cycle steps.4:00 PMRELAP5 Model of the High Flux Isotope Reactor with Low EnrichedFuel Thermal Flux ProfilesJ. Banfield, B. Mervin, S. Hart, J. Ritchie, S. Walker, A. Ruggles, G.I. MaldonadoDepartment of Nuclear Engineering, University of Tennessee Knoxville, Knoxville, TNThe High Flux Isotope Reactor (HFIR) currently uses highly enriched uranium (HEU)fabricated into involute-shaped fuel plates. It is desired that HFIR be able to use lowenriched uranium (LEU) fuel while preserving the current performance capability for itsdiverse missions in material irradiation studies, isotope production, and the use of neutronbeam lines for basic research. Preliminary neutronics and depletion simulations ofHFIR with LEU fuel have arrived to feasible fuel loadings that maintain the neutronicsperformance of the reactor. This article illustrates preliminary models developed forthe analysis of the thermal-hydraulic characteristics of the LEU core to ensure safeoperation of the reactor. The beginning of life (BOL) LEU thermal flux profile has beenmodeled in RELAP5 to facilitate steady state simulation of the core cooling, and ofanticipated and unanticipated transients. Steady state results are presented to validatethe new thermal power profile inputs. A power ramp, slow depressurization at the outlet,and flow coast down transients are also evaluated.43

44Chair: Ron Ellis (ORNL)PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 4:25 PM - Ballrom E-GPAPER 8An Analytical Solution for the Consideration of the Effect ofAdjacent Fuel Assemblies; Comparison of Rectangular andHexagonal StructuresB. Merk and U. RohdeHelmholtz-Zentrum Dresden-Rossendorf, Institute of Safety Research, Dresden, GermanyA new analytical method is described to deal with the Leakage Environmental Effect.The method is based on the analytical solution of the two-group diffusion equation fortwo adjacent fuel assemblies. The quality of the results for this highly efficient methodis demonstrated for square fuel assemblies. In additional tests the transferability ofthe concept to hexagonal VVER-440-type fuel assemblies is shown and a comparisonbetween the results for rectangular and hexagonal assemblies is given.PAPER 37Optimization of Irradiation Conditions for 177Lu Productionat the LVR-15 Research ReactorZdena Lahodová, Ladislav Viererbl, Vít Klupák (1), Jiří Šrank (2)1) Research Centre Řež Ltd., Řež, Czech Republic, 2) Nuclear Physics Institute of the Academy of Sciences,Řež, Czech RepublicThe use of lutetium in medicine has been increasing over the last few years. The 177Luradionuclide is commercially available for research and test purposes as a diagnosticand radiotherapy agent in the treatment of several malignant tumours. The yield of177Lu from the 176Lu(n,γ)177Lu nuclear reaction depends significantly on the thermalneutron fluence rate. The capture cross-sections of both reaction 176Lu(n,γ)177Luand reaction 177Lu(n,γ)178Lu are very high. Therefore a burn-up of target and productnuclides should be taken into account when calculating 177Lu activity. The maximumirradiation time, when the activity of the 177Lu radionuclide begins to decline,was found for different fluence rates. Two vertical irradiation channels at the LVR-15nuclear research reactor were compared in order to choose the channel with betterirradiation conditions, such as a higher thermal neutron fluence rate in the irradiationvolume. In this experiment, lutetium was irradiated in a titanium capsule. The influenceof the Ti capsule on the neutron spectrum was monitored using activation detectors.The choice of detectors was based on requirements for irradiation time and accuratedetermination of thermal neutrons. The following activation detectors were selected formeasurement of the neutron spectrum: Ti, Fe, Ni, Co, Ag and W.PAPER 52Design of a Proteus Lattice Representative of a Burnt andFresh Fuel Interface at Power Conditions in Light Water ReactorsMathieu Hursin and Gregory PerretPaul Scherrer Institut (PSI), Villigen, SwitzerlandThe research program LIFE (Large-scale Irradiated Fuel Experiment) between PSIand swissnuclear has been started in 2006 to study the interaction between large setsof burnt and fresh fuel pins in conditions representative of power light water reactors.Reactor physics parameters such as flux ratios and reaction rate distributions (235Uand 238U fissions and 238U capture) are calculated to estimate an appropriate arrangementof burnt and fresh fuel pins within the central element of the test zone ofthe zero-power research reactor PROTEUS. The arrangement should minimize thenumber of burnt fuel pins to ease fuel handling and reduce costs, whilst guaranteeingthat the neutron spectrum in both burnt and fresh fuel regions and at their interface isrepresentative of a large uniform array of burnt and fresh pins in the same moderationconditions. First results are encouraging, showing that the burnt/fresh fuel interfaceis well represented with a 6×6 bundle of burnt pins. The second part of the projectinvolves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric andoptimization studies on the PROTEUS lattice by varying its pitch (P) and fraction ofD2O in moderator (FD2O) to be as representative as possible of a power light waterreactor core at hot full power conditions at beginning of cycle (BOC). The parameters Pand FD2O that best represent a PWR at BOC are 1.36 cm and 5% respectively.PAPER 68Initiation of Persistent Fission Chains in the Fast Burst ReactorCALIBANNicolas Authier, Benoît Richard, Pascal Grivot, Pierre Casoli (1), PhilippeHumbert (2)1) Commissariat à l’Energie Atomique et aux Energies Alternatives, Is-sur-Tille, France. 2) Commissariatà l’Energie Atomique et aux Energies Alternatives, Arpajon Cedex, FrancePoster SessionWe provide in this article, experimental data of initiation of persistent fission chains obtainedat different supercritical states, using the Fast Burst Reactor CALIBAN. In manypast papers, theory has been compared mostly to initiation experiments at varioussuperprompt critical states, whereas very few experimental data has been publishedin delayed supercritical states. To fill the lack of data, we have conducted three campaignson the reactor at reactivities far below 0.7$ which was one of the rare loweststate ever published on a similar assembly[2][1]. We give a justification of the use ofthe gamma function to fit experimental results of the temporal distributions of waitingtimes and compare experiments with numerical simulations obtained with a zero-Dpunctual Monte Carlo code.PAPER 77On Fast Reactor Kinetics StudiesE.F. Seleznev, A.A. Belov (1), I.P. Matveenko, A.M. Zhukov, K.F. Raskach(2)1) Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE), Russia. 2) Institute for Physicsand Power Engineering (IPPE), RussiaThe results and the program of fast reactor core time and space kinetics experimentsperformed and planned to be performed at the IPPE critical facility is presented. TheTIMER code was taken as computation support of the experimental work, which allowstransient equations to be solved in 3-D geometry with multi-group diffusion approximation.The number of delayed neutron groups varies from 6 to 8. The code implementsthe solution of both transient neutron transfer problems: a direct one, where neutronflux density and its derivatives, such as reactor power, etc, are determined at each timestep, and an inverse one for the point kinetics equation form, where such a parameteras reactivity is determined with a well-known reactor power time variation function.PAPER 82Conceptual Design of Thorium-Fuelled Mitrailleuse Accelerator-DrivenSubcritical Reactor Using D-Be Neutron SourceYuji Kokubo (1), Takashi Kamei (2)1) Quan Japan Company Limited, Chuo-ku, Kobe, Hyogo, Japan. 2) Research Institute for Applied Sciences,Sakyo-ku, Kyoto-shi, Kyoto, JapanA distributed accelerator is a charged-particle accelerator that uses a new accelerationmethod based on repeated electrostatic acceleration. This method offers outstandingbenefits not possible with the conventional radio-frequency acceleration method,including: (1) high acceleration efficiency, (2) large acceleration current, and (3) lowerfailure rate made possible by a fully solid-state acceleration field generation circuit. A‘Mitrailleuse Accelerator’ is a product we have conceived to optimize this distributedaccelerator technology for use with a high-strength neutron source. We have completedthe conceptual design of a Mitrailleuse Accelerator and of a thorium-fuelledsubcritical reactor driven by a Mitrailleuse Accelerator. This paper presents the conceptualdesign details and approach to implementing the subcritical reactor core. Wewill spend the next year or so on detailed design work, and then will start work ondeveloping a prototype for demonstration. If there are no obstacles in setting up adevelopment organization, we expect to finish verifying the prototype’s performanceby the third quarter of 2015.PAPER 88Application of Wavelet Scaling Function Expansion Continuous-EnergyResonance Calculation Method to MOX Fuel ProblemYang Weiyan (1), Wu Hongchun and Cao Liangzhi (2)1) Shanghai Nuclear Engineering Research and Design Institute , Shanghai, China and Department ofNuclear Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi, China. 2) Department of Nuclear Engineering,Xi’an Jiaotong University, Xi’an, Shaanxi, ChinaMore and more MOX fuels are used in all over the world in the past several decades.Compared with UO2 fuel, it contains some new features. For example, the neutronspectrum is harder and more resonance interference effects within the resonance energyrange are introduced because of more resonant nuclides contained in the MOXfuel. In this paper, the wavelets scaling function expansion method is applied to studythe resonance behavior of plutonium isotopes within MOX fuel. Wavelets scaling functionexpansion continuous-energy self-shielding method is developed recently. It hasbeen validated and verified by comparison to Monte Carlo calculations. In this method,the continuous-energy cross-sections are utilized within resonance energy, whichmeans that it’s capable to solve problems with serious resonance interference effectswithout iteration calculations. Therefore, this method adapts to treat the MOX fuelresonance calculation problem natively. Furthermore, plutonium isotopes have fierceoscillations of total cross-section within thermal energy range, especially for 240Puand 242Pu. To take thermal resonance effect of plutonium isotopes into considerationthe wavelet scaling function expansion continuous-energy resonance calculation codeWAVERESON is enhanced by applying the free gas scattering kernel to obtain thecontinuous-energy scattering source within thermal energy range (2.1eV to 4.0eV)contrasting against the resonance energy range in which the elastic scattering kernelis utilized. Finally, all of the calculation results of WAVERESON are compared withMCNP calculation.

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 4:25 PM - Ballrom E-GPoster SessionPAPER 119Benchmark of Atucha-2 PHWR RELAP5-3D Control Rod Modelby Monte Carlo MCNP5 Core CalculationM. Pecchia, F. D’Auria (1), O. Mazzantini (2)1) San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Pisa, Italy. 2) NucleoelectricaArgentina Societad Anonima (NA-SA), Buenos Aires, ArgentinaAtucha-2 is a Siemens-designed PHWR reactor under construction in the Republicof Argentina. Its geometrical complexity and peculiarities require the adoption of advancedMonte Carlo codes for performing realistic neutronic simulations. Thereforecore models of Atucha-2 PHWR were developed using MCNP5. In this work a methodologywas set up to collect the flux in the hexagonal mesh by which the Atucha-2 coreis represented. The scope of this activity is to evaluate the effect of obliquely insertedcontrol rod on neutron flux in order to validate the RELAP5-3D©/NESTLE three dimensionalneutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPIfor performing selected transients of Chapter 15 FSAR of Atucha-2.PAPER 131Control Rod Reactivity Measurement by Rod-Drop Method ata Fast Critical AssemblyLingli Song, Yanpeng Yin, Xuan Lian, Chun ZhengInstitute of Nuclear Physics and Chemistry in CAEP, Mianyang, Sichuan, P. R. ChinaRod-drop experiments were carried out to estimate the reactivity of the control rod ofa fast critical assembly operated by CAEP. Two power monitor systems were used toobtain the power level and integration method was used to process the data. Threeexperiments were performed. The experimental results of the reactivity from the twopower monitor systems were consistent and showed a reasonable range of reactivitycompared to results from positive period method.PAPER 136Metastable Phases Determination Of U-2.5Zr-7.5Nb and U-3.0Zr-9.0Nb Alloys by Rietveld MethodRafael Witter Dias Pais, Ana Maria Matildes dos Santos, Fernando SoaresLameiras, Natália Mattar Cantagalli, Raphael Gomes de Paula and WilmarBarbosa FerrazCentro de Desenvolvimento da Tecnologia Nuclear, CDTN – CNEN, Belo Horizonte, MG, BrazilThe Rietveld refinement has been employed for study of metastable phase of alloysU-2.5Zr-7.5Nb(wt%) and U-3Zr-9Nb(wt%). The ingots of both alloys were producedin vacuum induction furnace at temperature of about 1500°C followed by cooling toroom temperature. The samples with 2.5 cm in diameter and 0.3 cm of thickness washomogenized at 1000°C/16 hours and treated isothermally at (i) 600°C for 0.5, 3 and24 hours and (ii) 300°C for 4 minutes, 20 minutes and 17.5 hours. At the end of eachtreatment the samples were water quenched. Data from X-ray diffraction were collectedat room temperature with a Rigaku diffractometer D\Max-RAPID radiation Cukusing steps of 0.02° (2θ) with scan angle in the range of 20-80° (2θ). The full diffractionpattern was analyzed by the Rietveld method using the GSAS program. The resultshows that the non-resolved appearance of the XRD patterns added to the proximityof the Bragg reflections of the transition phase makes the refinement of alloys a challengingtask.PAPER 140Coupling Procedure for Transuranus and KTF CodesJavier Jiménez, Samuel Alglave (1) and Maria Avramova (1,2)1) Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology (INR), Eggenstein-Leopoldshafen,Germany. 2) Department of Mechanical and Nuclear Engineering, The PennsylvaniaState University, PA, USAThe nuclear industry aims to ensure safe and economic operation of each single fuelrod introduced in the reactor core. This goal is even more challenging nowadays dueto the current strategy of going for higher burn-up (fuel cycles of 18 or 24 months)and longer residence time. In order to achieve that goal, fuel modeling is the key topredict the fuel rod behavior and lifetime under thermal and pressure loads, corrosionand irradiation. In this context, fuel performance codes, such as TRANSURANUS, areused to improve the fuel rod design. The modeling capabilities of the above mentionedtools can be significantly improved if they are coupled with a thermal-hydraulic code inorder to have a better description of the flow conditions within the rod bundle. For LWRapplications, a good representation of the two phase flow within the fuel assembly isnecessary in order to have a best estimate calculation of the heat transfer inside thebundle. In this paper we present the coupling methodology of TRANSURANUS withKTF (Karlsruher Twophase Flow subchannel code) as well as selected results of thecoupling proof of principle.PAPER 156Eastern Europe Research Reactor Initiative Nuclear Educationand Training Courses – Current Activities and FutureChallangesLuka Snoj (1), Lubomir Sklenka, Jan Rataj (2), Helmuth Böck (3)1) Jožef Stefan Institute, Ljubljana, Slovenia. 2) Department of Nuclear Reactor, Czech Technical Universityin Prague, Prague, Czech Republic. 3) Vienna University of Technology/Atominstitut, Vienna, AustriaThe Eastern Europe Research Reactor Initiative was established in January 2008 toenhance cooperation between the Research Reactors in Eastern Europe. It coversthree areas of research reactor utilisation: irradiation of materials and fuel, radioisotopeproduction, neutron beam experiments, education and training. In the field of educationand training an EERRI training course was developed. The training programmehas been elaborated with the purpose to assist IAEA Member States, which considerbuilding a research reactor (RR) as a first step to develop nuclear competence andinfrastructure in the Country. The major strength of the reactor is utilisation of threedifferent research reactors and a lot of practical exercises. Due to high level of adaptability,the course can be tailored to specific needs of institutions with limited or noaccess to research reactors.PAPER 1643-D Transient Anaylsis Of Pebble-Bed HTGR by TORT-TD/AT-TICA3DA. Seubert and A. Sureda (1), J. Lapins, M. Buck (2), J. Bader (2,3) and E.Laurien (2)1) Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH Forschungszentrum, Garching, Germany.2) Institut für Kernenergetik und Energiesysteme (IKE), Universität Stuttgart Pfaffenwaldring 31, Stuttgart,Germany. 3) EnBW Kernkraft GmbH, Kernkraftwerk Philippsburg, Rheinschanzinsel, Philippsburg,GermanyAs most of the acceptance criteria are local core parameters, application of transient3-D fine mesh neutron transport and thermal hydraulics coupled codes is mandatoryfor best estimate evaluations of safety margins. This also applies to high-temperaturegas cooled reactors (HTGR). Application of 3-D fine-mesh transient transport codesusing few energy groups coupled with 3-D thermal hydraulics codes becomes feasiblein view of increasing computing power. This paper describes the discrete ordinatesbased coupled code system TORT-TD/ATTICA3D that has recently been extended bya fine-mesh diffusion solver. Based on transient analyses for the PBMR-400 design,the transport/diffusion capabilities are demonstrated and 3-D local flux and power redistributioneffects during a partial control rod withdrawal are shown.PAPER 172Application of GRS Method to Evaluation of Uncertainties ofCalculation Parameters of Perspective Sodium-Cooled FastReactorAnton Peregudov, Olga Andrianova, Kirill Raskach and Anatoly TsibulyaInstitute for Physics and Power Engineering, Obninsk, RussiaA number of recent studies have been devoted to the estimation of errors of reactorcalculation parameters by the GRS (Generation Random Sampled) method. Thismethod is based on direct sampling input data resulting in formation of random setsof input parameters which are used for multiple calculations. Once these calculationsare performed, statistical processing of the calculation results is carried out to determinethe mean value and the variance of each calculation parameter of interest.In our study this method is used for estimation of errors of calculation parameters(Keff, power density, dose rate) of a perspective sodium-cooled fast reactor. Neutrontransport calculations were performed by the nodal diffusion code TRIGEX and MonteCarlo code ММК.PAPER 177An Example of Neutronic Penalizations in Reactivity TransientAnalysis Using 3D Coupled Chain HemeraF. Dubois, B. Normand, A. SargeniInstitut de Radioprotection et de sûreté Nucléaire, Reactor Safety Division, Fontenay-aux-Rose Cedex,FranceHEMERA (Highly Evolutionary Methods for Extensive Reactor Analyses), is a fullycoupled 3D computational chain developed jointly by IRSN and CEA. It is composedof CRONOS2 (core neutronics, cross sections library from APOLLO2), FLICA4 (corethermal-hydraulics) and the system code CATHARE. Multi-level and multi-dimensionalmodels are developed to account for neutronics, core thermal-hydraulics, fuel thermalanalysis and system thermal-hydraulics, dedicated to best-estimate, conservative simulationsand sensitivity analysis. In IRSN, the HEMERA chain is widely used to studyseveral types of reactivity accidents and for sensitivity studies. Just as an example ofthe HEMERA possibilities, we present here two types of neutronic penalizations andtheir impact on a power transient due to a REA (Rod Ejection Accident): in the first one,45

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 4:25 PM - Ballrom E-GPoster Sessionwe studied a burn-up distribution modification and in the second one, a delayedneutronfraction modification. Both modifications are applied to the whole core or localizedin a few assemblies. Results show that it is possible to use global or local changes but1) in case of burn-up modification, the total core power can increase when assemblypeak power decrease so, care has to be taken if the goal is to maximize a local powerpeak and 2) for delayed-neutron fraction, a local modification can have the same effectas the one on the whole core, provided that it is large enough.PAPER 178A Trigonal Nodal SP3 Method With Mesh Refinement Capabilities—Development and VerificationS. Duerigen, Y. Bilodid, E. Fridman, S. MittagHelmholtz-Zentrum Dresden-Rossendorf e.V.,Reactor Safety Division,Dresden, GermanyThe neutronics model of the nodal reactor dynamics code DYN3D developed for 3Danalyses of steady states and transients in Light-Water Reactors has been extendedby a simplified P3 (SP3) neutron transport option – to overcome the limitations of thediffusion approach at regions with significant anisotropy effects. To provide a methodbeing applicable to reactors with hexagonal fuel assemblies and to furthermore allowflexible mesh refinement, the nodal SP3 method has been developed on the basis ofa flux expansion in triangular-z geometry. In this paper, the derivation of the trigonalSP3 method is presented in a condensed form and a verification of the methodologyon quasi-pin level is performed by means of two single-assembly test examples. Thecorresponding pin-wise few-group cross sections were obtained by the deterministiclattice code HELIOS. The power distributions were calculated using both the trigonalDYN3D diffusion and SP3 solver and compared to the HELIOS reference solutions.Close to regions with non-negligible flux gradients, e.g., caused by the presence of astrong absorbing material, the power distribution calculated by DYN3D-SP3 shows asignificant improvement in comparison to the diffusion method.PAPER 180Neutron Noise Calculations in a Hexagonal Geometry andComparison with Analytical SolutionsH.N. Tran and C. DemazièreDepartment of Applied Physics, Division of Nuclear Engineering, Chalmers University of Technology,Gothenburg, SwedenThis paper presents the development of a neutronic and kinetic solver for hexagonalgeometries. The tool is developed based on the diffusion theory with multi-energygroups and multi-groups of delayed neutron precursors allowing the solutions of forwardand adjoint problems of static and dynamic states, and is applicable to boththermal and fast systems with hexagonal geometries. In the dynamic problems, thesmall stationary fluctuations of macroscopic cross sections are considered as noisesources, and then the induced first order noise is calculated fully in the frequencydomain. Numerical algorithms for solving the static and noise equations are implementedwith a spatial discretization based on finite differences and a power iterativesolution. A coarse mesh finite difference method has been adopted for speeding upthe convergence. Since no other numerical tool could calculate frequency-dependentnoise in hexagonal geometry, validation calculations have been performed and benchmarkedto analytical solutions based on a 2-D homogeneous system with two-energygroups and one-group of delayed neutron precursor, in which point-like perturbationsof thermal absorption cross section at central and non-central positions are consideredas noise sources.PAPER 208Benchmark Calculations on the Phase II Problem of UncertaintyAnalyses for Criticality Safety AssessmentGil Soo Lee, Jaejun Lee, Gwan-Young Kim, and Sweng-Woong WooKorea Institute of Nuclear Safety, Daejeon, KoreaThe phase II benchmark problem of expert group UACSA includes a configuration of aPWR fuel storage rack and focuses on the uncertainty of criticality from manufacturingtolerance of design parameters such as fuel enrichment, density, diameter, thicknessof neutron absorber and structural material, and so on. It provides probability densityfunctions for each design parameter. In this paper, upper limits of k-eff of 95%/95%tolerance with two methods are calculated by sampling design parameters using givenprobability distributions and compared with the result from traditional approach.PAPER 218Sensitivity of Posterior Parameter Correlations and ResponseParameter Correlations to Prior Response UncertaintyJ.J. WagschalRacah Institute of Physics, Hebrew University of Jerusalem, Jerusalem, IsraelThe two parameters one response model of the generalized linear least squaresparameter adjustment methodology was used in order to analyze the procedure ofimproving cross sections agreement with integral experimental results. Two extremeadjustment cases were considered. One case is using extremely accurate measuredintegral responses, such as keff of Godiva or Jezebel for instance. The other one ismodifying only one parameter, for instance the modification of - in ENDF/B-VII, in orderto improve agreement of calculated responses with corresponding integral measurementsresults. In both cases new posterior parameters correlations are generated.The conclusion is that it is not advised to use integral measurements information in theevaluation process of a general purpose nuclear data file.PAPER 226An Improved Energy-Collapsing Method for Core-ReflectorModelization in SFR Core Calculations Using the Paris PlatformJ-F. Vidal, P. Archier, A. Calloo, Ph. Jacquet, J. Tommasi (1), R. Le Tellier(2)1) CEA, DEN, DER/SPRC/LEPh, Saint-Paul-lez-Durance, France. 2) CEA, DEN, DTN/STRI/LMA, Saint-Paul-lez-Durance, FranceIn the framework of the ASTRID project, sodium cooled fast reactor studies are conductedat CEA in compliance with GEN IV reactors criteria, particularly for safety requirements.An improved safety requires better calculation tools to obtain accuratereactivity effects (especially sodium void effect) and power map distributions. The currentcalculation route lies on the JEFF3.1.1 library and the classical two-step approachperformed with the ECCO module of the ERANOS code system at the assembly leveland the Sn SNATCH solver - implemented within the PARIS platform - at the core level.33-group cross sections used by SNATCH are collapsed from 1968-group self-shieldedcross-section with a specific flux-current weighting. Recent studies have shownthat this collapsing is non-conservative when dealing with core-reflector interface andcan lead to reactivity discrepancies larger than 500 pcm in the case of a steel reflector.Such a discrepancy is due to the flux anisotropy at the interface, which is not taken intoaccount when cross sections are obtained from separate fuel and reflector assemblycalculations. A new approach is proposed in this paper. It consists in separating theself-shielding and the flux calculations. The first one is still performed with ECCO onseparate patterns. The second one is done with SNATCH on a 1D traverse, representativeof the core-reflector interface. An improved collapsing method using angular fluxmoments is then carried out to collapse the cross sections onto the 33-group structure.In the case of a simplified ZONA2B 2D homogeneous benchmark, results in terms ofkeff and power map are strongly improved for a small increase of the computing time.PAPER 240A Verification Regime for the Spatial Discretization of the SNTransport EquationsSebastian Schunert and Yousry AzmyNorth Carolina State University, Dept. of Nuclear Engineering, Raleigh, NCThe order-of-accuracy test in conjunction with the method of manufactured solutions isthe current state of the art in computer code verification. In this work we investigate theapplication of a verification procedure including the order-of-accuracy test on a genericSN transport solver that implements the AHOTN spatial discretization. Different typesof semantic errors, e.g. removal of a line of code or changing a single character, areintroduced randomly into the previously verified SN code and the proposed verificationprocedure is used to identify the coding mistakes (if possible) and classify them.Itemized by error type we record the stage of the verification procedure where theerror is detected and report the frequency with which the errors are correctly identifiedat various stages of the verification. Errors that remain undetected by the verificationprocedure are further scrutinized to determine the reason why the introduced codingmistake eluded the verification procedure. The result of this work is that the verificationprocedure based on an order-of-accuracy test finds almost all detectable coding mistakesbut rarely, 1:44% of the time, and under certain circumstances can fail.PAPER 247PWR Core and Spent Fuel Pool Analysis using SCALE andNESTLEJ. Evan Murphy and G. Ivan Maldonado (1), Robert St. Clair and David Orr(2)1) Department of Nuclear Engineering, University of Tennessee, Knoxville, TN. 2) Duke Energy, Charlotte,NCThe SCALE nuclear analysis code system [SCALE, 2011], developed and maintainedat Oak Ridge National Laboratory (ORNL) is widely recognized as high quality softwarefor analyzing nuclear systems. The SCALE code system is composed of severalvalidated computer codes and methods with standard control sequences, such as theTRITON/NEWT lattice physics sequence, which supplies dependable and accurateanalyses for industry, regulators, and academia. Although TRITON generates energycollapsedand space-homogenized few group cross sections, SCALE does not includea full-core nodal neutron diffusion simulation module within. However, in the past fewyears, the open-source NESTLE core simulator [NESTLE, 2003], originally developed46

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 4:25 PM - Ballrom E-GPoster Sessionat North Carolina State University (NCSU), has been updated and upgraded via collaborationbetween ORNL and the University of Tennessee (UT), so it now has a growinglyseamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. Thisstudy presents the methodology used to couple lattice physics data between TRITONand NESTLE in order to perform a three-dimensional full-core analysis employing a“real-life” Duke Energy PWR as the test bed. The focus for this step was to comparethe key parameters of core reactivity and radial power distribution versus plant data.Following the core analysis, following a three cycle burn, a spent fuel pool analysiswas done using information generated from NESTLE for the discharged bundles andwas compared to Duke Energy spent fuel pool models. The KENO control modulefrom SCALE was employed for this latter stage of the project.PAPER 249Processing of U-2.5Zr-7.5Nb and U-3Zr-9Nb Alloys by SinteringProcessAna Maria Matildes dos Santos, Wilmar Barbosa Ferraz, Fernando SoaresLameiras, and Thiago de Oliveira MazzeuCentro de Desenvolvimento da Tecnologia Nuclear, CDTN–CNEN, Belo Horizonte, MG-BrazilTo minimize the risk of nuclear proliferation, there is worldwide interest in reducingfuel enrichment of research and test reactors. To achieve this objective while still guaranteeingcriticality and cycle length requirements, there is need of developing highdensity uranium metallic fuels. Alloying elements such as Zr, Nb and Mo are added touranium to improve fuel performance in reactors. In this context, the Centro de Desenvolvimentoda Tecnologia Nuclear (CDTN) is developing the U-2.5Zr-7.5Nb and U-3Zr-9Nb (weight %) alloys by the innovative process of sintering that utilizes raw materialsin the form of powders. The powders were pressed at 400 MPa and then sinteredunder a vacuum of about 1x10-4 Torr at temperatures ranging from 1050o to 1500oC.The densities of the alloys were measured geometrically and by hydrostatic methodand the phases identified by X ray diffraction (XRD). The microstructures of the pelletswere observed by scanning electron microscopy (SEM) and the alloying elementswere analyzed by energy dispersive X-ray spectroscopy (EDS). The results obtainedshowed the fuel density to slightly increase with the sintering temperature. The highestdensity achieved was approximately 80% of theoretical density. It was observed in thepellets a superficial oxide layer formed during the sintering process.PAPER 263Improved Computational Neutronics Methods and ValidationProtocols for The Advanced Test ReactorDavid W. Nigg, Joseph W. Nielsen, Benjamin M. Chase, Ronnie K. Murray,Kevin A. Steuhm, Troy UnruhIdaho National Laboratory, Idaho Falls, IDThe Idaho National Laboratory (INL) is in the process of updating the various reactorphysics modeling and simulation tools used to support operation and safety assuranceof the Advanced Test Reactor (ATR). Key accomplishments so far have encompassedboth computational as well as experimental work. A new suite of stochastic and deterministictransport theory based reactor physics codes and their supporting nucleardata libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extendedimplementation of MCNP5) has been installed at the INL. Corresponding models ofthe ATR and ATRC are now operational with all five codes, demonstrating the basicfeasibility of the new code packages for their intended purposes. On the experimentalside of the project, new hardware was fabricated, measurement protocols were finalized,and the first four of six planned physics code validation experiments based onneutron activation spectrometry have been conducted at the ATRC facility. Data analysisfor the first three experiments, focused on characterization of the neutron spectrumin one of the ATR flux traps, has been completed. The six experiments will ultimatelyform the basis for flexible and repeatable ATR physics code validation protocols thatare consistent with applicable national standards.PAPER 271Tripoli-4 Criticality Calculations for MOX Fuelled SNEAK 7AAnd 7B Fast Critical AssembliesYi-Kang LeeCommissariat à l’Energie Atomique et aux Energies Alternatives, CEA-Saclay, Gif sur Yvette Cedex,FranceA prototype Generation IV fast neutron reactor is under design and development inFrance. The MOX fuel will be introduced into this self-generating core in order todemonstrate low net plutonium production. To support the TRIPOLI-4 Monte Carlotransport code in criticality calculations of fast reactors, the effective delayed neutronfraction beff estimation and the Probability Tables (PT) option to treat the unresolvedresonance region of cross-sections are two essentials. In this study, TRIPOLI-4 calculationshave been made using current nuclear data libraries JEFF-3.1.1 and ENDF/B-VII.0 to benchmark the reactor physics parameters of the MOX fuelled SNEAK 7A and7B fast critical assemblies. TRIPOLI-4 calculated Keff and beff of the homogeneousR-Z models and the 3D multi-cell models have been validated against the measuredones. The impact of the PT option on Keff is 340 + 10 pcm for SNEAK 7A core and410 + 12 pcm for 7B. Four-group spectra and energy spectral indices, f8/f5, f9/f5, andc8/f5 in the two SNEAK cores have also been calculated with the TRIPOLI-4 meshtally. Calculated spectrum-hardening index f8/f5 is 0.0418 for SNEAK 7A and 0.0315for 7B. From this study the SNEAK 3D models have been verified for the next revisionof IRPhE (International Handbook of Evaluated Reactor Physics Benchmark Experiments).PAPER 275Trade-off Study on the Power Capacity of a Prototype SFR inKoreaMin-Ho Baek, Sang-Ji Kim, Jaewoon Yoo and In-Ho Bae1) Labratoire de Physique Subatomique et de Cosmologie, CNRS-IN2P3/UJF/INPG, Grenoble, France.Korea Atomic Energy Research Institute, Daejeon, KoreaThe major roles of a prototype SFR are to provide irradiation test capability for the fueland structure materials, and to obtain operational experiences of systems. Due to acompromise between the irradiation capability and construction costs, the power levelshould be properly determined. In this paper, a trade-off study on the power level ofthe prototype SFR was performed from a neutronics viewpoint. To select candidatecores, the parametric study of pin diameters was estimated using 20 wt.% uraniumfuel. The candidate cores of different power levels, 125 MWt, 250 MWt, 400 MWt, and500 MWt, were compared with the 1500 MWt reference core. The resulting core performanceand economic efficiency indices became insensitive to the power at about400 ~ 500 MWt and sharply deteriorated at about 125 ~ 250 MWt with decreasing coresizes. Fuel management scheme, TRU core performance comparing with uraniumcore, and sodium void reactivity were also evaluated with increasing power levels. It isfound that increasing the number of batches showed higher burnup performance andeconomic efficiency. However, increasing the cycle length showed the trends in lowereconomic efficiency. Irradiation performance of TRU and enriched TRU cores wasimproved about 20 % and 50 %, respectively. The maximum sodium void reactivity of5.2$ was confirmed less than the design limit of 7.5$. As a result, the power capacityof the prototype SFR should not be less than 250MWt and would be appropriate at ~500 MWt considering the performance and economic efficiency.PAPER 278Research of a Boundary Condition Quantifiable CorrectionMethod in the Assembly HomogenizationPeng Liang-hui Liu Zhi-hong Zhao Jing (1), Li Wen-huai (2)1) Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing, China. 2) China NuclearPower Technology Research Institute, Shenzhen, ChinaThe methods and codes currently used in assembly homogenization calculation mostlyadopt the reflection boundary conditions. The influences of real boundary conditionson the assembly homo genized parameters were analyzed. They were summarizedinto four quantifiable effects, and then the mathematical expressions could be got bylinearization hypothesis. Through the calculation of a test model, it had been foundthat the result was close to transport calculation result when considering four boundaryquantifiable effects. This method would greatly improve the precision of a core designcode which using the assembly homogenization methods, but without much increaseof the computing time.PAPER 291Hybrid Method of Deterministic and Probabilistic Approachesfor Multigroup Neutron Transport ProblemDeokjung LeeUlsan National Institute of Science and Technology, Ulju-gun, Ulsan, KoreaA hybrid method of deterministic and probabilistic methods is proposed to solve Boltzmanntransport equation. The new method uses a deterministic method, Method ofCharacteristics (MOC), for the fast and thermal neutron energy ranges and a probabilisticmethod, Monte Carlo (MC), for the intermediate resonance energy range. Thehybrid method, in case of continuous energy problem, will be able to take advantage offast MOC calculation and accurate resonance selfshielding treatment of MC method.As a proof of principle, this paper presents the hybrid methodology applied to a multigroupform of Boltzmann transport equation and confirms that the hybrid method canproduce consistent results with MC and MOC methods.47

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 4:25 PM - Ballrom E-GPoster SessionPAPER 313Design Stress Evaluation Based on Strain-Rate SensitivityAnalysis for Nickel Alloys Used in the Very High TemperatureNuclear SystemKun Mo (1), Hsiao-Ming Tung, Xiang Chen, Yang Zhao and James F. Stubbins(2)Reactor Design and Fuel Management Research Center, China Nuclear Power Technology ResearchInstitute, Shenzhen, China. 2) Department of Nuclear, Plasma and Radiological EngineeringUniversity of Illinois at Urbana-Champaign, Urbana, ILBoth Alloy 617 and Alloy 230 have been considered the most promising structuralmaterials for the Very High Temperature Reactor (VHTR). In this study, mechanicalproperties of both alloys were examined by performing tensile tests at three differentstrain rates and at temperatures up to 1000ºC. This range covers time-dependent(plasticity) to time-independent (creep) deformations. Strain-rate sensitivity analysisfor each alloy was conducted in order to approximate the long-term flow stresses.The strain-rate sensitivities for the 0.2% flow stress were found to be temperatureindependent (m ≈ 0) at temperatures ranging from room temperature to 700ºC due todynamic strain aging. At elevated temperatures (800-1000ºC), the strain-rate sensitivitysignificantly increased (m > 0.1). Compared to Alloy 617, Alloy 230 displayed higherstrain-rate sensitivities at high temperatures. This leads to a lower estimated long-termflow stresses. Results of this analysis were used to evaluate current American Societyof Mechanical Engineers (ASME) allowable design limits. According to the comparisonwith the estimated flow stresses, the allowable design stresses in ASME B&PV Codefor either alloy did not provide adequate degradation estimation for the possible longtermservice life in VHTR. However, rupture stresses for Alloy 617, developed in ASMEcode case N-47-28, can generally satisfy the safety margin estimated in the studyfollowing the strain-rate sensitivity analysis. Nevertheless, additional material developmentstudies might be required, since the design parameters for rupture stresses areconstrained such that current VHTR conceptual designs cannot satisfy the limits.PAPER 314Numerical Simulation of Reactivity Measurements in VVER-1000 ReactorA. Popykin, O Kavun, S. Shevchenko and R. ShevchenkoScientific and Engineering Center for Nuclear and Radiation Safety, Moscow, RussiaReactivity is one of the most used and important concepts in physics and nuclear reactorcalculations carried out for the safety analysis. Currently, design calculation of RussianVVER reactors are carried out with modern coupled time-dependent neutronicand heat hydraulic codes. They allow to perform numerical simulation of reactivitymeasurements. However, point kinetic model used for simulation of large reactivityinsertion leads to some issues. The paper discusses the numerical simulation of reactivitymeasurement, shows that the reactivity obtained from the steady state solutiondoes not always correspond to the measured value. Comparison of scram systemreactivity worth calculated and measured during physical start-up of unit 3, KalininNPP is presented.PAPER 317Application of Fully Ceramic Micro-Encapsulated Fuel forTransuranic Waste Recycling In PWRsCole Gentry and Ivan Maldonado (1), Andrew Godfrey, Kurt Terrani, andJess Gehin (2)1) Department of Nuclear Engineering, University of Tennessee Knoxville, Knoxville, TN. 2) Oak RidgeNational Laboratory, Oak Ridge, TNPresented is an investigation of the utilization of Tristructural-Isotropic (TRISO) particle-basedfuel designs for the recycling of transuranic (TRU) wastes in typical Westinghousefour-loop pressurized water reactors (PWRs). Though numerous studies haveevaluated the recycling of TRU in light water reactors (LWRs), this work differentiatesitself by employing TRU-loaded TRISO particles embedded within a SiC matrix andformed into pellets that can be loaded into standard 17×17 fuel element cladding. Thisapproach provides the capability of TRU recycling and, by virtue of the TRISO particledesign, will allow for greater burnup (i.e., removal of the need for UO2 mixing) andimproved fuel reliability. In this study, a variety of assembly layouts and core loadingpatterns were analyzed to demonstrate the feasibility of TRU-loaded TRISO fuel. Theassembly and core design herein reported are a work in progress, so they still requiresome fine-tuning to further flatten power peaks; however, the progress achieved thusfar strongly supports the conclusion that with further rod/assembly/core loading andplacement optimization, TRU-loaded TRISO fuel and core designs that are capableof balancing TRU production and destruction can be designed within the standardconstraints for thermal and reactivity performance in PWRs.PAPER 318Convergence Analysis of a CMFD Method Based on GeneralizedEquivalence TheoryYunlin Xu (1), T. Downar (2)1) Argonne National Laboratory, Argonne, IL. 2) Department of Nuclear Engineering and RadiologicalSciences, University of Michigan, Ann Arbor, MICMFD acceleration methods have been successful in reducing the computationalburden for steady-state and transient reactor calculations. However, recent work ona complex coupled code BWR ATWS event has exposed possible issues with thestability of the CMFD method when standard CMFD methods are used. During thesimulation of the ATWS boron injection event in the BWR, the PARCS code failed toconverge with the existing CMFD method. A new CMFD method based on generalizedequivalence theory was developed and the PARCS solution converged for thesame ATWS event. This paper presents the new method and a detailed analytic andnumerical convergence analysis. The results show that the new CMFD converges forall possible cross sections combinations anticipated in Light Water Reactor simulationand unlike existing CMFD methods, it is very robust even when the initial guess is farfrom final true solution.PAPER 321Effect of Steam Generator Configuration in a Loss of the RHRDuring Midloop Operation at PKL FacilityJ.F. Villanueva, S. Carlos, S. Martorell and F. SánchezDpto. Ingeniería Química y Nuclear, Universitat Politècnica de València, Valencia, SpainThe loss of the residual heat removal system in mid-loop conditions may occur with anonnegligible contribution to the plant risk, so the analysis of the accidental sequencesand the actions to mitigate the accident are of great interest in shutdown conditions.In order to plan the appropriate measures to mitigate the accident is necessary tounderstand the thermal-hydraulic processes following the loss of the residual heatremoval system during shutdown. Thus, transients of this kind have been simulatedusing best-estimate codes in different integral test facilities and compared with experimentaldata obtained in different facilities. In PKL (Primärkreislauf- Versuchsanlage,primary coolant loop test facility) test facility different series of experiments have beenundertaken to analyze the plant response in shutdown. In this context, the E3 and F2series consist of analyzing the loss of the residual heat removal system with a reducedinventory in the primary system. In particular, the experiments were developed to investigatethe influence of the steam generators secondary side configuration on theplant response, what involves the consideration of different number of steam generatorsfilled with water and ready for activation, on the heat transfer mechanisms insidethe steam generators U-tubes. This work presents the results of such experimentscalculated using, RELAP5/Mod 3.3.PAPER 322High-Fidelity Multiphysics Simulation of BWR Assembly withCoupled TORT-TD/CTFJ. Magedanz (1), Y. Perin (2), M. Avramova (1), A. Pautz, F. Puente-Espel,A. Seubert, A. Sureda, K. Velkov, W. Zwermann (2)1) Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, UniversityPark, PA, USA. 2) Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Garching, GermanyThis paper describes the application of the coupled codes TORT-TD and CTF to thepin-by-pin modeling of a BWR fuel assembly with thermal-hydraulic feedback. TORT-TD, developed at GRS, is a time-dependent three dimensional discrete ordinatescode. CTF is the PSU’s improved version of the subchannel code COBRA-TF, whichuses a two-fluid, three-field model to represent two-phase flow with entrained droplets,and is commonly applied to evaluate LWR safety margins. The coupled codes systemTORT-TD/CTF, already applied to several PWR cases involving MOX, was adaptedfrom PWR to BWR applications. The purpose of the research described in this paper isto verify the coupling for modeling two-phase flow at the pin cell level. Using an ATRI-UM-10 assembly, the system’s steady-state capabilities were tested on two cases: onewithout control blade insertion and another with partially inserted blades. The influenceof the neutron absorber on local axial and radial parameters is presented. The descriptionof an inlet flow reduction transient is an example for the time-dependent capabilityof the coupled system.PAPER 324The Relative Variational Model: A Topological View of Matterand its Properties: Specific Heat and EnthalpyMarcio S. Dias, Vanderley de Vasconcelos, João Roberto L. Mattos (1),Elizabete Jordão (2)1) Center for Development of the Nuclear Technology – CDTN, National Commission for the NuclearEnergy – CNEN, Minas Gerais, Brazil. 2) Chemistry Engineering Dept., Campinas State University, FEQ/UNICAMP, São Paulo, BrazilFormal definitions of convergence, connectedness and continuity were established tocharacterize and describe the crystalline solid and its properties as a unified notion inthe topological space. The crystalline solid is a previously empty space that has beenfilled with atoms and phonons, i.e., the crystal is built with packages of matter and48

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 4:25 PM - Ballrom E-GPoster Sessionenergy in a regular and orderly repetitive pattern along three orthogonal dimensions ofthe space. The spatial occupation of the atom in the crystal structure is determined byits mean vibrational volume. Thus, the changes of volume and the changes of internalenergy are intrinsically linked. In fact, physical and material properties are the interdependentand bijective quantifications associated with variations of the internal energy.These properties are modeled by means of an intrinsic and invariable form function:the Relative Variational Model. In this paper, the Debye’s integral of the heat capacityat constant volume is analytically solved. The experimental data of the specific heatat constant pressure and the enthalpy variations are also analytically depicted by themodel in the temperature range of 0 K up to the melting point. The data reductionswere applied to the oxides Al2O3 and UO2.PAPER 347Verification of the Shift Monte Carlo Code with the C5G7 ReactorBenchmarkNicholas C. Sly, Brenden T. Mervin (1), Scott W. Mosher, Thomas M. Evans,John C. Wagner (2), and G. Ivan Maldonado (1)1) Department of Nuclear Engineering, University of Tennessee, Knoxville, TN. 2) Oak Ridge NationalLaboratory, Oak Ridge, TNShift is a new hybrid Monte Carlo/deterministic radiation transport code being developedat Oak Ridge National Laboratory. At its current stage of development, Shiftincludes a parallel Monte Carlo capability for simulating eigenvalue and fixed-sourcemultigroup transport problems. This paper focuses on recent efforts to verify Shift’sMonte Carlo component using the two-dimensional and three-dimensional C5G7 NEAbenchmark problems. Comparisons were made between the benchmark eigenvaluesand those output by the Shift code. In addition, mesh-based scalar flux tally resultsgenerated by Shift were compared to those obtained using MCNP5 on an identicalmodel and tally grid. The Shift-generated eigenvalues were within three standard deviationsof the benchmark and MCNP5-1.60 values in all cases. The flux tallies generatedby Shift were found to be in very good agreement with those from MCNP.PAPER 349Probabilistic Methods in a Study Of Trip SetpointsDale E. KaulitzTennessee Valley Authority, Chattanooga, TennesseeMost early vintage Boiling Water Reactors have a high head and high capacity HighPressure Coolant Injection (HPCI) pump to keep the core covered following a loss ofcoolant accident (LOCA). However, the protection afforded by the HPCI pump for mitigatinga LOCA introduces the potential that a spurious start of the HPCI pump couldoversupply the reactor vessel and lead to an automatic trip of the main turbine dueto high water level. A turbine trip and associated increase in moderator density couldchallenge the bases of fuel integrity operating limits. To prevent turbine trip during spuriousoperation of the HPCI pump, the reactor protection system includes instrumentationand logic to sense high water level and automatically trip the HPCI pump prior toreaching the turbine trip setpoint. This paper describes an analysis that was performedto determine if existing reactor vessel water level trip instrumentation, logic and setpointsresult in a high probability that the HPCI pump will trip prior to actuation of theturbine trip. Using nominal values for the initial water level and for the HPCI pump andturbine trip setpoints, and using the probability distribution functions for measurementuncertainty in these setpoints, a Monte Carlo simulation was employed to determineprobabilities of successfully tripping the HPCI pump prior to tripping of the turbine. Theresults of the analysis established that the existing setpoints, instrumentation and logicwould be expected to reliably prevent a trip of the main turbine.PAPER 352Assessment of Fission Product Yields Data Needs in NuclearReactor ApplicationsKilian Kern, Maarten Becker, Cornelis BroedersInstitut für Neutronenphysik und Reaktortechnik, KIT Campus Nord, Leopoldshafen, GermanyStudies on the build-up of fission products in fast reactors have been performed, withparticular emphasis on the effects related to the physics of the nuclear fission process.Fission product yields, which are required for burn-up calculations, depend on the protonand neutron number of the target nucleus as well as on the incident neutron energy.Evaluated nuclear data on fission product yields are available for all relevant targetnuclides in reactor applications. However, the description of their energy dependencein evaluated data is still rather rudimentary, which is due to the lack of experimentalfast fission data and reliable physical models. Additionally, physics studies of evaluatedJEFF-3.1.1 fission yields data have shown potential improvements, especially forvarious fast fission data sets of this evaluation. In recent years, important progress inthe understanding of the fission process has been made, and advanced model codesare currently being developed. This paper deals with the semi-empirical approach tothe description of the fission process, which is used in the GEF code being developedby K.-H. Schmidt and B. Jurado on behalf of the OECD Nuclear Energy Agency, andwith results from the corresponding author’s diploma thesis. An extended version ofthe GEF code, supporting the calculation of spectrum weighted fission product yields,has been developed. It has been applied to the calculation of fission product yields inthe fission rate spectra of a MOX fuelled sodium-cooled fast reactor. Important resultsare compared to JEFF-3.1.1 data and discussed in this paper.PAPER 354A Compact Breed and Burn Fast Reactor Using Spent NuclearFuel BlanketDonny Hartanto and Yonghee KimKorea Advanced Institute of Science and Technology (KAIST), Daejeon, KoreaA long-life breed-and-burn (B&B) type fast reactor has been investigated from theneutronics points of view. The B&B reactor has the capability to breed the fissile fuelsand use the bred fuel in situ in the same reactor. In this work, feasibility of a compactsodium-cooled B&B fast reactor using spent nuclear fuel as blanket material has beenstudied. In order to derive a compact B&B fast reactor, a tight fuel lattice and relativelylarge fuel pin are used to achieve high fuel volume fraction. The core is initially loadedwith an LEU (Low Enriched Uranium) fuel and a metallic fuel is used in the core. TheMonte Carlo depletion has been performed for the core to see the long-term behaviorof the B&B reactor. Several important parameters such as reactivity coefficients,delayed neutron fraction, prompt neutron generation lifetime, fission power, and fastneutron fluence, are analyzed through Monte Carlo reactor analysis. Evolution of thecore fuel composition is also analyzed as a function of burnup. Although the long-lifesmall B&B fast reactor is found to be feasible from the neutronics point of view, it ischaracterized to have several challenging technical issues including a very high fastneutron fluence of the structural materials.PAPER 357Source Term Characterization for SNM PIT Storage FacilitiesMichael Chin, Jessica Paul, and Glenn SjodenNuclear and Radiological Engineering Program, George W. Woodruff School of Mechanical Engineering,Georgia Institute of Technology, Atlanta, GAIn order to properly design a mobile system to validate and verify the presence of specialnuclear materials for non-proliferation and safeguards applications, accurate modelingof source materials is imperative. In this work, models were developed for usein design assessments based on an AL-R8 SNM standardized container specificationto determine the radioactive signatures for both highly enriched uranium (HEU) andweapons plutonium (WGPu) special nuclear materials (SNM) housed in the containers.Intrinsic gamma boundary leakage currents were evaluated for this system, performedusing 3D fixed-source deterministic SN photon transport (PENTRAN) as well as withstochastic Monte Carlo methods (MCNP5). Group-dependent leakage radiation termswere calculated at two “source box” interfaces within the models, one directly surroundingthe SNM source, and one immediately surrounding the canister. Analysisshowed good agreement between the two models for energy groups of interest, basedon a 24 group gamma library established for HEU and WGPu gamma signatures ofinterest. Intrinsic and neutron induced gamma leakage was determined using MonteCarlo calculations, and the combined gamma signatures were then treated as a netgamma leakage to be used in subsequent photon transport calculations. Neutron leakagebased on the BUGLE-96 47 group structure was determined using Monte Carlocalculations for the WGPu canisters. These results will be used to evaluate the sourceterm from stored nuclear materials and augment our efforts to design a detection systemto validate the presence of these materials for safeguards purposes.PAPER 360Neutron/Gamma Coupled Library Generation and GammaTransport Calculation With KARMA 1.2Ser Gi Hong (1), Kang-Seog Kim, Jin Young Cho and Kyung Hoon Lee (2)1) Department of Nuclear Engineering, Kyung Hee University, Gyeonggi-do, Korea. 2) Korea AtomicEnergy Research Institute, Daejon, KoreaKAERI has developed a lattice transport calculation code KARMA and its multi-groupcross section library generation system. Recently, the multi-group cross section librarygeneration system has included a gamma cross section generation capability andKARMA also has been improved to include a gamma transport calculation module.This paper addresses the multi-group gamma cross section generation capability forthe KARMA 1.2 code and the preliminary test results of the KARMA 1.2 gamma transportcalculations. The gamma transport calculation with KARMA 1.2 gives the gammaflux, gamma smeared power, and gamma energy deposition distributions. The resultsof the KARMA gamma calculations were compared with those of HELIOS and theyshowed that KARMA 1.2 gives reasonable gamma transport calculation results.49

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 4:25 PM - Ballrom E-GPoster SessionPAPER 363Fast Reactor 3D Core and Burnup Analysis Using VESTANicholas Luciano, Jacob Shamblin, Ivan MaldonadoNuclear Engineering Department, The University of Tennessee, Knoxville, TNBurnup analyses using the VESTA code have been performed on a MOX-fuelled fastreactor model as specified by an IAEA computational benchmark. VESTA is a relativelynew code that has been used for burnup credit calculations and thermal reactormodels, but not typically for fast reactor applications. The detailed input and resultsof the IAEA benchmark provides an opportunity to gauge the use of VESTA in a fastreactor application. VESTA employs an ultra-fine multi-group binning approach thataccelerates Monte Carlo burnup calculations. Using VESTA to compute the end ofcycle (EOC) power fractions by enrichment zone showed agreement with the publishedvalues within 5%. When comparing the ultra-fine multi-group binning approachto the tally-based approach, EOC isotopic masses also agree within 5%. Using theultra-fine multi-group binning approach, we obtain a wall-time speedup factor of 35when compared to the tally-based approach for computing a ke eigenvalue with burnupproblem. The authors conclude the use of VESTA’s ultra-fine multi-group binningapproach with Monte Carlo transport performs accurate depletion calculations for thisfast reactor benchmark.PAPER 364Validation of the U.S. NRC NGNP Evaluation Model with theHTTRThomas Saller, Volkan Seker, and Tom DownarDepartment of Nuclear Engineering, University of MichiganThe High Temperature Test Reactor (HTTR) was modeled with TRITON/PARCS. Traditionallight water reactor (LWR) homogenization methods rely on the short mean freepaths of neutrons in LWR. In gas-cooled, graphite-moderated reactors like the HTTRneutrons have much longer mean free paths and penetrate further into neighboringassemblies than in LWRs. Because of this, conventional lattice calculations with asingle assembly may not be valid. In addition to difficulties caused by the longer meanfree paths, the HTTR presents unique axial and radial heterogeneities that require additionalmodifications to the single assembly homogenization method. To handle thesechallenges, the homogenization domain is decreased while the computational domainis increased. Instead of homogenizing a single hexagonal fuel assembly, the assemblyis split into six triangles on the radial plane and five blocks axially in order to accountfor the placement of burnable poisons. Furthermore, the radial domain is increasedbeyond a single fuel assembly to account for spectrum effects from neighboring fuel,reflector, and control rod assemblies. A series of five two-dimensional cases, eachcloser to the full core, were calculated to evaluate the effectiveness of the homogenizationmethod and cross-sections.PAPER 380Neutronics Studies of Uranium-Based Fully Ceramic Micro-Encapsulated Fuel For PWRsNathan Michael George and Ivan Maldonado (1), Kurt Terrani, Andrew Godfrey,and Jess Gehin (2)1) Department of Nuclear Engineering University of Tennessee Knoxville Knoxville, TN. 2) Oak RidgeNational Laboratory, Oak Ridge, TNThis study evaluates the core neutronics and fuel cycle characteristics using uraniumbasedfully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor(PWR). Specific PWR assembly designs with FCM fuel have been developed, whichby virtue of their TRISO particle- based elements are expected to achieve higher fuelburnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package,developed and maintained at ORNL, was the primary software used to model theassembly designs. Analysis was performed using the SCALE double- heterogeneous(DH) fuel modeling capabilities; however, the Reactivity-Equivalent Physical Transformation(RPT) method was used for lattice calculations due to the long run timesassociated with the SCALE DH capability. In order to understand the impact on reactivityand reactor operating cycle length, a parametric study was performed by varyingTRISO particle design features, such as kernel diameter, coating layer thicknesses,and packing fraction. Also, other features such as the selection of matrix material (SiC,zirconium) and fuel rod dimensions were studied. After evaluating different uraniumbasedfuels, the higher compound density of uranium mononitride (UN) proved to befavorable, as the parametric studies showed that the FCM particle fuel design willneed roughly 12% additional fissile material in comparison to that of a standard UO2rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCMfuel designs evaluated maintain acceptable design features in the areas of fuel lifetimeand temperature coefficients of reactivity, as well as pin cell and assembly peakingfactors.PAPER 394The Method of Characteristics for 2-D Multigroup and PointwiseTransport Calculations in SCALE/CENTRMKang Seog Kim and Mark L. WilliamsOak Ridge National Laboratory, Oak Ridge, TN, USASCALE 6 computes problem-dependent multigroup (MG) cross sections through acombination of the conventional Bondarenko shielding-factor method and a deterministicpointwise (PW) transport calculation of the fine-structure spectra in the resolvedresonance and thermal energy ranges. The PW calculation is performed by the CEN-TRM code using a 1-D cylindrical Wigner-Seitz model with the white boundary conditioninstead of the real rectangular cell shape to represent a lattice unit cell. Thepointwise fluxes computed by CENTRM are not exact because a 1-D model is used forthe transport calculation, which introduces discrepancies in the MG self-shielded crosssections, resulting in some deviation in the eigenvalue. In order to solve this problem,the method of characteristics (MOC) has been applied to enable the CENTRM PWtransport calculation for a 2-D square pin cell. The computation results show that thenew BONAMI/CENTRM-MOC procedure produces very precise self-shielded crosssections compared to MCNP reaction rates.PAPER 401Sensitivity and Uncertainty in the Effective Delayed NeutronFraction (ßeff)I. I. KodeliJozef Stefan Institute, Ljubljana, SloveniaPrecise knowledge of effective delayed neutron fraction (βeff) and of the correspondinguncertainty is important for reactor safety analysis. The interest in developing themethodology for estimating the uncertainty in βeff was expressed in the scope of theUAM project of the OECD/NEA. A novel approach for the calculation of the nucleardata sensitivity and uncertainty of the effective delayed neutron fraction is proposed,based on the linear perturbation theory. The method allows the detailed analysis ofcomponents of βeff uncertainty. The procedure was implemented in the SUSD3D sensitivityand uncertainty code applied to several fast neutron benchmark experimentsfrom the ICSBEP and IRPhE databases. According to the JENDL-4 covariance matricesand taking into account the uncertainty in the cross sections and in the prompt anddelayed fission spectra the total uncertainty in βeff was found to be of the order of ~2to ~3.5 % for the studied fast experiments.PAPER 404The Relative Variational Model - A Topological View of Matterand its Properties: Thermal ExpansionMarcio S. Dias, Vanderley de Vasconcelos, João Roberto L. Mattos (1),Elizabete Jordão (2)1) Center for Development of the Nuclear Technology – CDTN, National Commission for the NuclearEnergy – CNEN, Minas Gerais, Brazil. 2) Chemistry Engineering Dept., Campinas State University, FEQ/UNICAMP, São Paulo, BrazilFormal definitions of convergence, connectedness and continuity were established tocharacterize and describe the crystalline solid and its properties as a unified notion inthe topological space. The crystalline solid is a previously empty space that has beenfilled with atoms and phonons, i.e., the crystal is built with packages of matter andenergy in a regular and orderly repetitive pattern along three orthogonal dimensions ofthe space. The spatial occupation of the atom in the crystal structure is determined byits mean vibrational volume. Thus, the changes of volume and the changes of internalenergy are intrinsically linked. In fact, physical and material properties are the interdependentand bijective quantifications associated with variations of the internal energy.These properties are modeled by means of an intrinsic and invariable form function:the Relative Variational Model. In this paper, the experimental data of the thermalexpansion for the oxides Al2O3 and UO2 were analytically depicted by means of thismodel in the temperature range of 0 K up to the melting point.PAPER 405The Relative Variational Model - A Topological View of Matterand its Properties: UO2±x DensityMarcio S. Dias, Vanderley de Vasconcelos, João Roberto L.Mattos, FernandoS. Lameiras (1), Elizabete Jordão (2)1) Center for Development of the Nuclear Technology – CDTN, National Commission for the NuclearEnergy – CNEN, Minas Gerais, Brazil. 2) Chemistry Engineering Dept., Campinas State University, FEQ/UNICAMP, São Paulo, BrazilFormal definitions of convergence, connectedness and continuity were established tocharacterize and describe the crystalline solid and its properties as a unified notion inthe topological space. The crystalline solid is a previously empty space that has beenfilled with atoms and phonons, i.e., the crystal is built with packages of matter andenergy in a regular and orderly repetitive pattern along three orthogonal dimensions of50

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 4:25 PM - Ballrom E-GPoster Sessionthe space. The spatial occupation of the atom in the crystal structure is determined byits mean vibrational volume. Thus, the changes of volume and the changes of internalenergy are intrinsically linked. In fact, physical and material properties are the interdependentand bijective quantifications associated with variations of the internal energy.These properties are modeled by means of an intrinsic and invariable form function:the Relative Variational Model. In this paper, the lattice parameter and specific mass ofUO2±x are depicted in dependence on the variations of the stoichiometry in the bandof − 0.016 < x

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationTuesday April 17, 2012 - 4:25 PM - Ballrom E-GPoster Sessionexperience demonstrates that TQRM’s results are accurate and real-time, the architectureis stable, and it could be extended and maintained conveniently for any otherRisk-Informed Application.PAPER 429Development and Validation of Instantaneous Risk Model inNuclear Power Plant’s Risk MonitorJiaqun Wang, Yazhou Li, Fang Wang, Jin Wang, Liqin Hu (1,2)1) Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences. 2) School of NuclearScience and Technology, University of Science and Technology of China, Anhui, ChinaThe instantaneous risk model is the fundament of calculation and analysis in a riskmonitor. This study focused on the development and validation of an instantaneousrisk model. Therefore the principles converting from the baseline risk model to theinstantaneous risk model were studied and separated trains’ failure modes modelingmethod was developed. The development and validation process in an operatingnuclear power plant’s risk monitor were also introduced. Correctness of instantaneousrisk model and rationality of converting method were demonstrated by comparisonwith the result of baseline risk model.52

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 8:00 AM - 301 A12B - Sensitivity & Uncertainty AnalysisSession Chair: Tatiana Ivanova (IRSN), Hany Abdel-Khalik (NCSU)8:00 AMSensitivity Analysis of MONJU Using Eranos with JENDL-4.0P. Tamagno (1), W. F. G. Van Rooijen, T. Takeda (2),M. Konomura (3)1) Institut National des Sciences et Techniques Nucléaires, Centre CEA de Saclay, Gif-sur-Yvette CedexFRANCE. 2) Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho, JAPAN. 3)Japan Atomic Energy Agency, FBR Plant Engineering Center, Shiraki JAPANThis paper deals with sensitivity analysis using JENDL-4.0 nuclear data applied to theMonju reactor. In 2010 the Japan Atomic Energy Agency – JAEA – released a new setof nuclear data: JENDL-4.0. This new evaluation is expected to contain improved dataon actinides and covariance matrices. Covariance matrices are a key point in quantificationof uncertainties due to basic nuclear data. For sensitivity analysis, the well-establishedERANOS [1] code was chosen because of its integrated modules that allowusers to perform a sensitivity analysis of complex reactor geometries. A JENDL-4.0cross-section library is not available for ERANOS. Therefore a cross-section libraryhad to be made from the original nuclear data set, available as ENDF formatted files.This is achieved by using the following codes: NJOY, CALENDF, MERGE and GECCOin order to create a library for the ECCO cell code (part of ERANOS). In order tomake sure of the accuracy of the new ECCO library, two benchmark experiments havebeen analyzed: the MZA and MZB cores of the MOZART program measured at theZEBRA facility in the UK. These were chosen due to their similarity to the Monju core.Using the JENDL-4.0 ECCO library we have analyzed the criticality of Monju duringthe restart in 2010. We have obtained good agreement with the measured criticality.Perturbation calculations have been performed between JENDL-3.3 and JENDL-4.0based models. The isotopes 239Pu, 238U, 241Am and 241Pu account for a major partof observed differences.8:25 AMNuclear Data Uncertainty Propagation for Neutronic Key Parametersof CEA’S SFR V2B and CFV Sodium Fast ReactorDesignsP. Archier, L. Buiron, C De Saint Jean and N. Dos SantosCEA, DEN, DER/SPRC, Saint Paul-lez-Durance, FranceThis paper presents a nuclear data uncertainty propagation analysis for two CEA’sSodium-cooled Fast Reactor designs: the SFR V2b and CFV cores. The nuclear datacovariance matrices are provided by the DER/SPRC/LEPh’s nuclear data team (seecompagnion paper) for several major isotopes. From the current status of this analysis,improvements on certain nuclear data reactions are highlighted as well as the needfor new specific integral experiments in order to meet the technological breakthroughsproposed by the CFV core.8:50 AMSensitivity and Uncertainty Analysis Applied to the JHR ReactivityPredictionO. Leray, C. Vaglio-Gaudard, J.P. Hudelot, A. Santamarina, G. Noguere (1)and J. Di-Salvo (2)Commissariat à l’Energie Atomique et aux Energies Alternatives 1) CEA, DEN, DER, SPRC, St Paul-Lez-Durance, France. 2) CEA, DEN, DER, SPEx, St Paul-Lez-Durance, FranceThe on-going AMMON program in EOLE reactor at CEA Cadarache (France) providesexperimental results to qualify the HORUS-3D/N neutronics calculation scheme usedfor the design and safety studies of the new Material Testing Jules Horowitz Reactor(JHR). This paper presents the determination of technological and nuclear datauncertainties on the core reactivity and the propagation of the latter from the AMMONexperiment to JHR. The technological uncertainty propagation was performed with adirect perturbation methodology using the 3D French stochastic code TRIPOLI4 and astatistical methodology using the 2D French deterministic code APOLLO2-MOC whichleads to a value of 289 pcm (1s). The Nuclear Data uncertainty propagation relies ona sensitivity study on the main isotopes and the use of a retroactive marginalizationmethod applied to the JEFF 3.1.1 27Al evaluation in order to obtain a realistic multigroupcovariance matrix associated with the considered evaluation. This nuclear datauncertainty propagation leads to a Keff uncertainty of 624 pcm for the JHR core and684 pcm for the AMMON reference configuration core. Finally, transposition and reductionof the prior uncertainty were made using the Representativity method whichdemonstrates the similarity of the AMMON experiment with JHR (the representativityfactor is 0.95). The final impact of JEFF 3.1.1 nuclear data on the Begin Of Life (BOL)JHR reactivity calculated by the HORUS-3D/N V4.0 is a bias of +216 pcm with an associatedposterior uncertainty of 304 pcm (1s).9:15 AMUncertainty Estimation of Core Safety Parameters UsingCross-Correlations of Covariance MatrixAkio Yamamoto, Yoshihiro Yasue, Tomohiro Endo (1), Yasuhiro Kodama,Yasunori Ohoka, Masahiro Tatsumi (2)1) Graduate School of Engineering, Nagoya University, Nagoya, Japan. 2) Nuclear Fuel Industries, Ltd.,Osaka, JapanAn uncertainty estimation method for core safety parameters, for which measurementvalues are not obtained, is proposed. We empirically recognize the correlations amongthe prediction errors among core safety parameters, e.g., a correlation between thecontrol rod worth and assembly relative power of corresponding position. Correlationsof uncertainties among core safety parameters are theoretically estimated using thecovariance of cross sections and sensitivity coefficients for core parameters. The estimatedcorrelations among core safety parameters are verified through the direct Monte-Carlosampling method. Once the correlation of uncertainties among core safetyparameters is known, we can estimate the uncertainty of a safety parameter for whichmeasurement value is not obtained. Furthermore, the correlations can be also used forthe reduction of uncertainties of core safety parameters.9:40 AMSCALE Code Validation for Prismatic High-Temperature Gas-Cooled ReactorsDan IlasOak Ridge National Laboratory, Oak Ridge, TN, USAUsing experimental data published in the International Handbook of Evaluated ReactorPhysics Benchmark Experiments for the fresh cold core of the High TemperatureEngineering Test Reactor, a comprehensive validation study has been carried out toassess the performance of the SCALE code system for analysis of high-temperaturegas-cooled reactor configurations. This paper describes part of the results of that effort.The studies performed included criticality evaluations for the full core and for theannular cores realized during the fuel loading, as well as calculations and comparisonsfor excess reactivity, shutdown margin, control rod worths, temperature coefficient ofreactivity, and reaction rate distributions. Comparisons of the SCALE results with bothexperimental values and MCNP-calculated values are presented. The comparisonsshow that the SCALE calculated results, obtained with both multigroup and continuousenergy cross sections, are in reasonable agreement with the experimental data andthe MCNP predictions.53

8:00 AMSetting the Stage for the Work Force of the Future –The Next Generation Safeguards Initiative’s University Curriculum DevelopmentThe panel will focus on addressing the human resource development component of the Department of Energy’s Next GenerationSafeguards Initiative (NGSI). Five universities in the east partnered with Oak Ridge National Laboratory to incorporatenuclear security concepts (nuclear nonproliferation and safeguards) into existing nuclear engineering courses. Universityprofessors on the panel will discuss how they accomplish the goal and address future plans.Panel Organizer: Bernie KirkPHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 8:00 AM - 301 B5A - Panel on Education in Reactor PhysicsSession Chair: B. Kirk (ORNL, ret)Panelists:Jason Hayward (University of Tennessee)Nolan Hertel (Georgia Tech)Sara Pozzi (University of Michigan)Steve Skutnik (North Carolina State University)9:15 AM“Responsible Science” and the ISTC Programs, SupportingCollaboration, Training and Education for Nuclear Nonproliferation,Security and Safety CultureL.V. TochenyISTC - International Science and Technology Center, Moscow, RussiaThe International Science and Technology Center (ISTC) is a unique international organization,established in Moscow in 1994 by Russia, USA, EU and Japan. Later Republicof Korea, Canada and several CIS countries acceded to ISTC. The basic ideabehind establishing the ISTC was to support non-proliferation of knowledge relatedto weapons of mass destruction by re-directing former Soviet weapons scientists topeaceful research thus preventing the drain of dual use knowledge and expertise fromRussia and other CIS countries. Presently, the ISTC has 39 member states (27 fromEU), representing the CIS, Europe, Asia, and North America. The Partner list includesover 200 organizations and leading industrial companies from all ISTC parties. Numerousscience and technology projects were realized with the ISTC support in areasranging from biotechnologies and environmental problems to all aspects of nuclearscience and technology. Many projects were focused on the development of innovativeconcepts and technologies in the nuclear field including many aspects of nuclearsafety. The new program “Responsible Science” is mainly focused on nonproliferationand security issues. The following activities are implementing: The pilot lectures/programs on “responsible science”, like courses on non-proliferation principles, arestarted in universities: MEPhI in Moscow, OIAtE- MEPhI in Obninsk and TPI in Tomskfor students and young experts. The program supports activities of think-tanks andother organizations dealing with nonproliferation activities and raising public awarenessin these matters. Related research and training programs to obtain and build-upknow-how on the vital scientific and technical aspects of overall nonproliferation policiesunder conditions of “nuclear renaissance”, safeguards, export control, dual-usetechnology monitoring, etc., are supported as well.9:40 AMCollaborative Development of Estonian Nuclear Master’s ProgramAlan H. Tkaczyk, Arvo Kikas, Enn Realo, Marco Kirm, Madis Kiisk, KadriIsakar, Siiri Suursoo, Rein Koch, Eduard Feldbach, Aleksandr Lushchik,Kaido ReiveltInstitute of Physics, University of Tartu, Tartu, EstoniaIn 2009 Estonia approved the National Development Plan for the Energy Sector, includingthe nuclear energy option. This can be realized by construction of a nuclearpower plant (NPP) in Estonia or by participation in neighboring nuclear projects (e.g.,Lithuania and/or Finland). Either option requires the availability of competent personnel.It is necessary to prepare specialists with expertise in all aspects related to nuclearinfrastructure and to meet workforce needs (e.g. energy enterprises, public agencies,municipalities). Estonia’s leading institutions of higher education and research with thesupport of the European Social Fund have announced in this context a new nuclearmaster’s curriculum to be developed. The language of instruction will be English.54

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 8:00 AM - 301 C9B - Research Reactors & Spallation SourcesSession Chair: Irina Popova (ORNL/SNS)8:00 AMOverview of SNS Accelerator Shielding AnalysesI. Popova and F. X. Gallmeier, P. Ferguson, E. Iverson, Wei LuORNL/SNS, Oak Ridge, TNThe Spallation Neutron Source is an accelerator driven neutron scattering facility formaterials research. During all phases of SNS development, including design, construction,commissioning and operation, extensive neutronics work was performed inorder to provide adequate shielding, to assure safe facility operation from radiationprotection point of view, and to optimize performance of the accelerator and targetfacility. Presently, most of the shielding work is concentrated on the beam lines andinstrument enclosures to prepare for commissioning, safe operation and adequateradiation background in the future. Although the accelerator is built and in operationmode, there is extensive demand for shielding and activation analyses. It includes redesigningsome parts of the facility, facility upgrades, designing additional structures,storage and transport containers for accelerator structures taken out of service, andperforming radiation protection analyses and studies on residual dose rates inside theaccelerator.8:25 AMConceptual Design and Optimization for a Low-Power ADSwith a 70 MeV, 0.75 mA Proton BeamS. Frambati, L. Mansani (1), M. Osipenko, G. Ricco, M. Ripani, P. Saracco,C.M. Viberti (2)1) Ansaldo Nucleare S.p.A., Genova, (Italy). 2) INFN - sez. Genova and Dept. of Physics, Genoa University,Genova (Italy)In the framework of research on Generation IV reactors, it is very important to haveinfrastructures specifically dedicated to the study of fundamental kinetic and dynamicparameters of future fast-neutron reactors, a capability not provided by many presentlyavailable or planned zero-power prototypes. We propose the conceptual design of anADS with high safety standards, in order to be used as a training facility, but also agood flexibility so as to allow for a wide range of static an dynamic measurements andexperiments. A high safety standard is guaranteed by limiting the system power to lessthan 500 kW and the neutron multiplication coefficient keff to less than 0:95 (a limitingvalue for waste storage areas), by using pure Uranium fuel (no Plutonium) and byusing solid Lead as a diffuser. Lead has been chosen by considering this prototype tobe a useful step towards the design of future Lead Fast Reactors, as well as becauseit allows a harder neutron spectrum, which facilitates tests on actinides fission, andprovides meaningful data on the physics of this metal, which is one of the proposedcoolants for future Generation IV reactors. The system is intrinsically subcritical and itneeds an external neutron source to be sustained. Specific goal of the present workis to optimize design features of the core in such a way to meet previously declaredrequirements with the use of a commercially available accelerator in order to reducecosts and increase reliability. The conceptual design requires, in the “full configuration”,60 active elements, each made by a solid Lead matrix of dimensions 92 921200mm3, each containing 81 Uranium oxide fuel rods, enriched 20%w=o in 235U.Protons, coming from a continuous cyclotron of 70MeV in energy and 0:75mA in beamcurrent, are converted into neutrons by a Beryllium target. Cooling is provided by Heliumgas, transparent to neutron and not subject to activation.8:50 AMComparisons on Thin and Thick Neutron Target for Low EnergyProton BeamBin Zhong, Ganglin Yu, Xuewu Wang, and Kan WangDepartment of Engineering Physics, Tsinghua University, Beijing, ChinaAs the progress on accelerator physics and neutronics, the compact neutron sourcesdriven by low energy and high intensity beam are becoming extensively developedand researched all around the world. The neutron target of an accelerator driven neutronsource is one of the key components, and the stability of the neutron target affectthe operation and performance of the neutron facility. When a low energy proton isprojected to the beryllium target, the main reaction is the inelastic scattering betweenthe proton and extra-nuclear electrons. As the decreasing of proton energy, the rateof elastic scattering between proton and target nucleus begins to increase. When theenergy of proton is very low, the pickup charge reaction begins to appear. Focus on theproblems brought by high intensity proton beam such as proton implantation, radiationdamages, heat deposition and gas production, we performed sufficient numericalsimulations for both thin and thick target determined by proton range. The results showthat the critical problem for thick target is the proton implantation, causing the formingof bubbles and beryllium flaked in vacuum. The thin target sacrifices a little neutronyield, but avoid the proton stopped in target, and decrease the radiation damage andenergy deposition.9:15 AMExperimental Study on the Thorium-Loaded Accelerator-DrivenSystem at the Kyoto University Critical AssemblyCheol Ho Pyeon, Takahiro Yagi, Jae-Yong Lim and Tsuyoshi MisawaNuclear Science Engineering Division, Research Reactor Institute, Kyoto University, Osaka, JapanThe experimental study on the thorium-loaded accelerator-driven system (ADS) isconducted in the Kyoto University Critical Assembly (KUCA). The experiments arecarried out in both the critical and subcritical states for attaining the reaction rates ofthe thorium capture and fission reactions. In the critical system, the thorium plate irradiationexperiment is carried out for the thorium capture and fission reactions. Fromthe results of the measurements, the thorium fission reactions are obtained apparentlyin the critical system, and the C/E values of reaction rates show the accuracy of relativedifference of about 30%. In the ADS experiments with 14 MeV neutrons and 100MeV protons, the subcritical experiments are carried out in the thorium-loaded coresto obtain the capture reaction rates through the measurements of 115In(n, γ)116mInreactions. The results of the experiments reveal the difference between the reactionrate distributions for the change in not only the neutron spectrum but also the externalneutron source. The comparison between the measured and calculated reaction ratedistributions demonstrates a discrepancy of the accuracy of reaction rate analyses ofthorium capture reactions through the thorium-loaded ADS experiments with 14 MeVneutrons. Hereafter, kinetic experiments are planned to be carried out to deduce thedelayed neutron decay constants and subcriticality using the pulsed neutron method.9:40 AMModeling Filters for Formation of Mono-Energetic NeutronBeams in the Research Reactor IRT Mephi and Optimizationof Radiation Shielding for Liquid-Xenon DetectorS.V. Ivakhin and G.V.Tikhomirov, A.I. Bolozdynya, Yu.V. Efremenko, D.Yu.Akimov, V.N. StekhanovLaboratory of Experimental Nuclear Physics, National Research Nuclear University «MEPhI», Moscow,RUSSIAThe paper considers formation of mono-energetic neutron beams at the entrance ofexperimental channels in research reactors for various applications. The problem includesthe following steps: 1. Full-scale mathematical model of the research IRT ME-PhI was developed for numerical evaluations of neutron spectra and neutron spatialdistribution in the area of experimental channels. 2. Modeling of filters in the channelto shift neutron spectrum towards the required mono-energetic line was performed. 3.Some characteristics of neutron beams at the entrance of detector were evaluated.The filter materials were selected. The calculations were carried out with application ofthe computer code based on the high-precision Monte-Carlo code MCNP. As a result,mathematical model was created for the filter which is able to form mono-energetic (24keV) neutron beam. The study was carried out within the frames of the research projecton development of Russian emission detector with liquid noble gas to observe rareprocesses of neutrino scattering and particles of hypothetical dark matter in atomicnuclei.55

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 8:00 AM - 301 D7A - Fuel Cycle & Actinide ManagementSession Chair: Mick Mastilovic (TVA); Fausto Franceschini (Westinghouse)8:00 AMPWR Core Design, Neutronics Evaluation and Fuel CycleAnalysis for Thorium-Uranium Breeding RecycleGuangwen BI, Chanyun LIU, Shengyi SIShanghai Nuclear Engineering Research & Design Institute No. 29, Shanghai ChinaThis paper was focused on core design, neutronics evaluation and fuel cycle analysisfor Thorium- Uranium Breeding Recycle in current PWRs, without any major changeto the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-basedpellet. The fuel cycle analysis indicates that Thorium-Uranium BreedingRecycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR coreutilizing 17×17 fuel assemblies (FAs) was taken as the model core. Two mixed coreswere investigated respectively loaded with mixed reactor grade Plutonium- Thorium(PuThOX) FAs and mixed reactor grade 233U-Thorium (U3ThOX) FAs on the basis ofreference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixedcore consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade 233U extractedfrom burnt PuThOX fuel was used to fabrication of U3ThOX for starting Thorium- Uraniumbreeding recycle. In UOX/U3ThOX mixed core, the well designed U3ThOX FAswith 1.94w/o fissile uranium (mainly 233U) were located on the periphery of core as ablanket region. U3ThOX FAs remained in-core for 6 cycles with the discharged burnupachieving 28GWD/tHM. Compared with initially loading, the fissile material inventoryin U3ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuelassemblies were located in the inner of UOX/U3ThOX mixed core refueling with 64FAs at each cycle. The designed UOX/PuThOX and UOX/U3ThOX mixed core satisfiedrelated nuclear design criteria. The full core performance analyses have shownthat mixed core with PuThOX loading has similar impacts as MOX on several neutroniccharacteristic parameters, such as reduced differential boron worth, higher criticalboron concentration, more negative moderator temperature coefficient, reduced controlrod worth, reduced shutdown margin, etc.; while mixed core with U3ThOX loadingon the periphery of core has no visible impacts on neutronic characteristics comparedwith reference full UOX core. The fuel cycle analysis has shown that 233U monorecyclingwith U3ThOX fuel could save 13% of natural uranium resource comparedwith UOX once through fuel cycle, slightly more than that of Plutonium single-recyclingwith MOX fuel. If 233U multi-recycling with U3ThOX fuel is implemented, more naturaluranium resource would be saved.8:25 AMExperimental Validation of The DARWIN2.3 Package for FuelCycle ApplicationsL. San-Felice, R. Eschbach, P. Bourdot (1), A. Tsilanizara, T.D. Huynh (2),H. Ourly (3), J.F. Thro (4)1) CEA, 1DEN, DER, SPRC, CEA-Cadarache, Paul-Lez-Durance, France. 2 ) CEA 2DEN, DM2S, SER-MA, CEA-Saclay, Gif sur Yvette, France. 3) EDF, Clamart Cedex, France. 4) AREVA, Tour AREVA, Parisla Défense, FranceThe DARWIN package, developed by the CEA and its French partners (AREVA andEDF) provides the required parameters for fuel cycle applications: fuel inventory, decayheat, activity, neutron, γ, α, β sources and spectrum, radiotoxicity. This paperpresents the DARWIN2.3 experimental validation for fuel inventory and decay heatcalculations on Pressurized Water Reactor (PWR). In order to validate this code systemfor spent fuel inventory a large program has been undertaken, based on spentfuel chemical assays. This paper deals with the experimental validation of DARWIN2.3for the Pressurized Water Reactor (PWR) Uranium Oxide (UOX) and Mixed Oxide(MOX) fuel inventory calculation, focused on the isotopes involved in Burn-Up Credit(BUC) applications and decay heat computations. The calculation – experiment (C/E-1) discrepancies are calculated with the latest European evaluation file JEFF-3.1.1associated with the SHEM energy mesh. An overview of the tendencies is obtainedon a complete range of burnup from 10 to 85 GWd/t (10 to 60GWd/t for MOX fuel).The experimental validation of the DARWIN2.3 package for decay heat calculation isperformed using calorimetric measurements carried out at the Swedish Interim SpentFuel Storage Facility for Pressurized Water Reactor (PWR) assemblies, covering alarge burn-up (20 to 50 GWd/t) and cooling time range (10 to 30 years).8:50 AMA High Converter Concept for Fuel Management with BlanketFuel Assemblies in Boiling Water ReactorsN. Martínez-Francès, W. Timm, D. RoßbachAREVA, AREVA NP, Erlangen, GermanyStudies on the natural Uranium saving and waste reduction potential of a multipleplantBWR system were performed. The BWR High Converter system should enablea multiple recycling of MOX fuel in current BWR plants by introducing blanket fuelassemblies and burning Uranium and MOX fuel separately. The feasibility of Uraniumcores with blankets and full-MOX cores with Plutonium qualities as low as 40% werestudied. The power concentration due to blanket insertion is manageable with modernfuel and acceptable values for the thermal limits and reactivity coefficients were obtained.While challenges remain, full-MOX cores also complied with the main designcriteria. The combination of Uranium and Plutonium burners in appropriate proportionscould enable obtaining as much as 40% more energy out of Uranium ore. Moreover,a proper adjustment of blanket average stay and Plutonium qualities could lead to asystem with nearly no Plutonium left for final disposal. The achievement of such goalswith current light water technology makes the BWRHC concept an attractive option toimprove the fuel cycle until Gen-IV designs are mature.9:15 AMQualification of JEFF3.1.1 Library for High Conversion ReactorCalculations Using the ERASME/R ExperimentJ-F. Vidal, G. Noguere, Y. Pénéliau, A. SantamarinaCEA, DEN, DER/SPRC/LEPh, Saint-Paul-lez-Durance, FranceWith its low CO2 production, Nuclear Energy appears to be an efficient solution to theglobal warming due to green-house effect. However, current LWR reactors are pooruranium users and, pending the development of Fast Neutron Reactors, alternativeconcepts of PWR with higher conversion ratio (HCPWR) are being studied again atCEA, first studies dating from the middle 80’s. In these French designs, low moderationratio has been performed by tightening the lattice pitch, achieving a conversionratio of 0.8-0.9 with a MOX fuel coming from PWR UOX recycling. Theses HCPWRsare characterized by a harder neutron spectrum and the calculation uncertainties onthe fundamental neutronics parameters are increased by a factor 3 regarding a standardPWR lattice, due to the major contribution of the Plutonium isotopes and of theepithermal energy range to the reaction rates. In order to reduce these uncertainties,a 3-year experimental validation program called ERASME has been performed byCEA from 1984 to 1986 in the EOLE reactor. Monte Carlo analysis of the ERASME/Rexperiments with the Monte Carlo code TRIPOLI4 allowed the qualification of therecommended JEFF.3.1.1 library for major neutronics parameters. Keff of the MOXunder-moderated lattice is overpredicted by 440 ± 830 pcm (2s); the conversion ratio,indicator of the good use of uranium, is also slightly overpredicted : 2 % ± 4 % (2s) andthe same for B4C absorber rods worth and soluble boron worth, overpredicted by 2%, both in the 2 standard deviations range. The radial fission maps of heterogeneities(water-holes, B4C and fertile rods) are well reproduced: maximal (C-E)/E dispersion is1.3 %, maximal power peak error is 2.7 %. The void reactivity worth is the only parameterpoorly calculated with an overprediction of +12.4 % ± 1.5 %.9:40 AMEquilibrium Cycle Pin By Pin Transport Depletion CalculationsWith DecartBrendan Kochunas and Thomas Downar (1), Temitope Taiwo (2)1) Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MIUSA. 2) Argonne National Laboratory, Argonne, IL, USAAs the Advanced Fuel Cycle Initiative (AFCI) program has matured it has becomemore important to utilize more advanced simulation methods. The work reported herewas performed as part of the AFCI fellowship program to develop and demonstrate thecapability of performing high fidelity equilibrium cycle calculations. As part of the workhere, a new multi-cycle analysis capability was implemented in the DeCART codewhich included modifying the depletion modules to perform nuclide decay calculations,implementing an assembly shuffling pattern description, and modifying iterationschemes. During the work, stability issues were uncovered with respect to convergingsimultaneously the neutron flux, isotopics, and fluid density and temperature distributionsin 3-D. Relaxation factors were implemented which considerably improvedthe stability of the convergence. To demonstrate the capability two core designs wereutilized, a reference UOX core and a CORAIL core. Full core equilibrium cycle calculationswere performed on both cores and the discharge isotopics were compared.From this comparison it was noted that the improved modeling capability was notdrastically different in its prediction of the discharge isotopics when compared to 2-Dsingle assembly or 2-D core models. For fissile isotopes such as U-235, Pu-239, andPu-241 the relative differences were 1.91%, 1.88%, and 0.59%, respectively. Whilethis difference may not seem large it translates to mass differences on the order oftens of grams per assembly, which may be significant for the purposes of accountingof special nuclear material.56

Session Chair: Mike Dunn (ORNL)PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 8:00 AM - 301 E8:00 AMThe Importance of Covariance in Nuclear Data UncertaintyPropagation StudiesJames BensteadAWE Plc, Aldermaston, Berkshire, United KingdomA study has been undertaken to investigate what proportion of the uncertainty propagatedthrough plutonium critical assembly calculations is due to the covariances betweenthe fission cross section in different neutron energy groups. The uncertaintieson keff calculated show that the presence of covariances between the cross section indifferent neutron energy groups accounts for approximately 27-37% of the propagateduncertainty due to the plutonium fission cross section. This study also confirmed thevalidity of employing the sandwich equation, with associated sensitivity and covariancedata, instead of a Monte Carlo sampling approach to calculating uncertainties forlinearly varying systems.8:25 AMReactivity Impact of 16O Thermal Elastic-Scattering NuclearData for Some Numerical and Critical Benchmark SystemsKen S. Kozier and Dan Roubtsov (1), Arjan J.M. Plompen and Stefan Kopecky(2)1) AECL – Chalk River Laboratories, Chalk River, Ontario, Canada. 2) EC-JRC-Institute for ReferenceMaterials and Measurements, Geel, BelgiumThe thermal neutron-elastic-scattering cross-section data for 16O used in variousmodern evaluated-nuclear-data libraries were reviewed and found to be generallytoo high compared with the best available experimental measurements. Some of theproposed revisions to the ENDF/B-VII.0 16O data library and recent results from theTENDL system increase this discrepancy further. The reactivity impact of revising the16O data downward to be consistent with the best measurements was tested using theJENDL-3.3 16O cross-section values and was found to be very small in MCNP5 simulationsof the UO2 and reactor-recycle MOX-fuel cases of the ANS Doppler-defect numericalbenchmark. However, large reactivity differences of up to about 14 mk (1400pcm) were observed using 16O data files from several evaluated-nuclear-data librariesin MCNP5 simulations of the Los Alamos National Laboratory HEU heavy-water solutionthermal critical experiments, which were performed in the 1950‘s. The latter resultsuggests that new measurements using HEU in a heavy-water-moderated critical facility,such as the ZED-2 zero-power reactor at the Chalk River Laboratories, might helpto resolve the discrepancy between the 16O thermal elastic-scattering cross-sectionvalues and thereby reduce or better define its uncertainty, although additional assessmentwork would be needed to confirm this.8:50 AMTowards the Reanalysis of Void Coefficients Measurementsat Proteus for High Conversion Light Water Reactor LatticesMathieu Hursin, Oliver Koeberl and Gregory PerretPaul Scherrer Institut (PSI), Villigen PSI, SwitzerlandHigh Conversion Light Water Reactors (HCLWR) allows a better usage of fuel resourcesthanks to a higher breeding ratio than standard LWR. Their uses together withthe current fleet of LWR constitute a fuel cycle thoroughly studied in Japan and theUS today. However, one of the issues related to HCLWR is their void reactivity coefficient(VRC), which can be positive. Accurate predictions of void reactivity coefficient inHCLWR conditions and their comparisons with representative experiments are thereforerequired. In this paper an inter comparison of modern codes and cross-sectionlibraries is performed for a former Benchmark on Void Reactivity Effect in PWRs conductedby the OECD/NEA . It shows an overview of the k-inf values and their associatedVRC obtained for infinite lattice calculations with UO2 and highly enriched MOXfuel cells. The codes MCNPX2.5, TRIPOLI4.4 and CASMO-5 in conjunction with thelibraries ENDF/B-VI.8, -VII.0, JEF-2.2 and JEFF-3.1 are used. A non-negligible spreadof results for voided conditions is found for the high content MOX fuel. The spread ofeigenvalues for the moderated and voided UO2 fuel are about 200 pcm and 700 pcm,respectively. The standard deviation for the VRCs for the UO2 fuel is about 0.7% whilethe one for the MOX fuel is about 13%. This work shows that an appropriate treatmentof the unresolved resonance energy range is an important issue for the accurate determinationof the void reactivity effect for HCLWR. A comparison to experimental resultsis needed to resolve the presented discrepancies.11A - Nuclear Data9:15 AMEstimation of multi-group cross section covariances for235,238U, 239Pu, 241Am, 56Fe, 23Na and 27AlC. De Saint Jean, P. Archier, G. Noguere, O. Litaize, C. Vagliogaudard, D.Bernard And O. LerayCEA, DEN, DER, SPRC, Cadarache, F-13108 Saint-Paul-lez-Durance, France.This paper presents the methodology used to estimate multi-group covariances forsome major isotopes used in reactor physics. The starting point of this evaluation isthe modelling of the neutron induced reactions based on nuclear reaction models withparameters. These latest are the vectors of uncertainties as they are absorbing uncertaintiesand correlation arising from the confrontation of nuclear reaction model tomicroscopic experiment. These uncertainties are then propagated towards multi-groupcross sections. As major breakthroughs were then asked by nuclear reactor physiciststo assess proper uncertainties to be used in applications, a solution is proposed by theuse of integral experiment information at two different stages in the covariance estimation.In this paper, we will explain briefly the treatment of all type of uncertainties,including experimental ones (statistical and systematic) as well as those coming fromvalidation of nuclear data on dedicated integral experiment (nuclear data oriented).We will illustrate the use of this methodology with various isotopes such as 235,238U,239Pu, 241Am, 56Fe, 23Na and 27Al.9:40 AMPreliminary Resolved Resonance Region Evaluation of Copper-63from 0 to 300 kevVladimir Sobes and Benoit Forget (1), Luiz Leal and Klaus Guber (2)1) Department of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge,MA. 2) Oak Ridge National Laboratory, Oak Ridge, TNA new preliminary evaluation of Cu-63 was done in the energy region from 0 to 300keV extending the resolved resonance region of the previous, ENDF/B-VII.0, evaluationthree-fold. The new evaluation was based on three experimental transmissiondata sets; two measured at the Oak Ridge Electron Linear Accelerator (ORELA) andone from the Massachusetts Institute of Technology Nuclear Reactor (MITR). A total of275 new resonances were identified and a corresponding set of external resonanceswas approximated to mock up the external levels. The negative external levels (boundlevel) were modified to match the thermal cross section values. A preliminary benchmarkingcalculation was made using 11 ICSBEP benchmarks. This work is in supportof the DOE Nuclear Criticality Safety Program.57

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 10:20 AM - 301 A1D - Core Analysis MethodsSession Chair: Jesse Cheatem (TerraPower)10:20 AMImplementation and Validation of the Variational Nodal ExpansionMethod Core SimulatorSteven Mullet, Makoto Tsuiki and William Beere (1), Urban Sandberg andHenrik Nylén (2)1) Institute for Energy Technology, OECD Halden Reactor Project, Halden, Norway. 2) Ringhals AB,Väröbacka, SwedenThe variational nodal expansion method (VNEM) has been developed by the OECDHalden Reactor Project for the purpose of implementation into a full-core simulator.This version of the VNEM does not make use of the diffusion approximation, but insteadprovides a fast converged solution to the even-parity neutron transport equationin several minutes using nodal methods and variational techniques. This “method tosimulator” implementation has been completed and the VNEM-simulator is currentlyundergoing continued validation efforts to ensure that it can track more nuclear reactorcycles effectively. This paper presents the results of the most recent developmenton the 3D VNEM core simulator implementation and validation efforts. The method isbriefly described and data is presented and discussed for three cases for the initialcriticality tests at the beginning of life, hot stand-by state for the Ringhals-3 PWR. Alsopresented here are the results for one year of operation of the maiden cycle of this unit.The effective neutron multiplication factor, keff, and relative neutron detector readingsare compared showing excellent agreement.11:10 AMKrylov Subspace Iteration for Eigenvalue Response MatrixCalculationsJeremy A. Roberts and Benoit ForgetMassachusetts Institute of Technology, Cambridge, MARecent work has revisited the eigenvalue response matrix method as an approachfor reactor core analyses. In its most straightforward form, the method consists of atwo-level eigenproblem. An outer Picard iteration updates the k-eigenvalue, while theinner eigenproblem imposes current continuity between coarse meshes. In this paper,several eigensolvers are evaluated for this inner problem, using several 2-D diffusionbenchmarks as test cases. The results indicate both the explicitlyrestarted Arnoldi andthe Krylov-Schur methods are up to an order of magnitude more efficient than poweriteration. This increased efficiency makes the nested eigenvalue formulation more effectivethan the ILU-preconditioned Newton-Krylov formulation previously studied.11:35 AMDevelopment of Iteration Strategies for a Practical Implementationof a Higher Order Transverse Leakage ApproximationRian H. Prinsloo and Djordje I TomaševićNecsa, Pretoria,South Africa10:45 AMNodal Weighting Factor Method for EXCORE Fast NeutronFluence EvaluationRen-Tai ChiangAREVA NP INC., San Jose, CA, USAThe nodal weighting factor method is developed for evaluating excore fast neutronflux in a nuclear reactor by utilizing adjoint neutron flux, a fictitious unit detector crosssection for neutron energy above 1 or 0.1 MeV, the unit fission source, and relativeassembly nodal powers. The method determines each nodal weighting factor for excoreneutron fast flux evaluation by solving the steady-state adjoint neutron transportequation with a fictitious unit detector cross section for neutron energy above 1 or0.1 MeV as the adjoint source, by integrating the unit fission source with a typical fissionspectrum to the solved adjoint flux over all energies, all angles and given nodalvolume, and by dividing it with the sum of all nodal weighting factors, which is a normalizationfactor. Then, the fast neutron flux can be obtained by summing the variousrelative nodal powers times the corresponding nodal weighting factors of the adjacentsignificantly contributed peripheral assembly nodes and times a proper fast neutron attenuationcoefficient over an operating period. A generic set of nodal weighting factorscan be used to evaluate neutron fluence at the same location for similar core designand fuel cycles, but the set of nodal weighting factors needs to be recalibrated for atransition-fuel-cycle. This newly developed nodal weighting factor method should bea useful and simplified tool for evaluating fast neutron fluence at selected locations ofinterest in excore components of contemporary nuclear power reactors.Transverse integrated nodal diffusion methods currently represent the standard in fullcore neutronic simulation. The primary shortcoming in this approach is the utilizationof the quadratic transverse leakage approximation. This approach, although provento work well for typical LWR problems, is not consistent with the formulation of nodalmethods and can cause accuracy and convergence problems. In previous work, animproved, consistent quadratic leakage approximation was formulated, which derivedfrom the class of higher order nodal methods developed some years ago. In this papera number of iteration schemes are developed around this consistent quadratic leakageapproximation which yield accurate node average results in much improved calculationaltimes. The developed consistent leakage approximation is extended in this workvia a number of numerical schemes, the most promising of which results from utilizingthe consistent leakage approximation as a correction method to the standard quadraticleakage approximation. Numerical results are demonstrated on a set of benchmarkproblems, such as the 3D IAEA LWR and MOX C5 problems.58

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 10:20 AM - 301 B5B - Education in Reactor PhysicsSession Chair: C. Pyeon (Kyoto U); Sandra Dulla (PdiT)10:20 AMRole of Research Reactors in Training of NPP Personnel withSpecial Focus on Training Reactor VR-1Sklenka, L., Rataj, J., Frybort, J. and Huml, O.Department of Nuclear Reactors, Czech Technical University in Prague, Prague, Czech Republic11:10 AMReactor Physics Teaching and Research in the Swiss NuclearEngineering MasterR. ChawlaSwiss Federal Institute of Technology (EPFL), CH-1015 Lausanne, SwitzerlandResearch reactors play an important role in providing key personnel of nuclear powerplants a hands-on experience from operation and experiments at nuclear facilities.Training of NPP (Nuclear Power Plant) staff is usually deeply theoretical with an extensiveutilisation of simulators and computer visualisation. But a direct sensing ofthe reactor response to various actions can only improve the personnel awarenessof important aspects of reactor operation. Training Reactor VR-1 and its utilization fortraining of NPP operators and other professionals from Czech Republic and Slovakiais described. Typical experimental exercises and good practices in organization of atraining program are demonstrated.10:45 AMAdvances in Reactor Physics Education: Visualization of ReactorParametersLuka Snoj, Marjan Kromar, Gašper ŽerovnikJožef Stefan Institute, Ljubljana, SloveniaModern computer codes allow detailed neutron transport calculations. In combinationwith advanced 3D visualization software capable of treating large amounts of data inreal time they form a powerful tool that can be used as a convenient modern educationaltool for reactor operators, nuclear engineers, students and specialists involvedin reactor operation and design. Visualization is applicable not only in education andtraining, but also as a tool for fuel management, core analysis and irradiation planning.The paper treats the visualization of neutron transport in different moderators, neutronflux and power distributions in two nuclear reactors (TRIGA type research reactor anda typical PWR). The distributions are calculated with MCNP and CORD-2 computercodes and presented using Amira software.Since 2008, a Master of Science program in Nuclear Engineering (NE) has beenrunning in Switzerland, thanks to the combined efforts of the country’s key playersin nuclear teaching and research, viz. the Swiss Federal Institutes of Technology atLausanne (EPFL) and at Zurich (ETHZ), the Paul Scherrer Institute (PSI) at Villigenand the Swiss Nuclear Utilities (swissnuclear). The present paper, while outlining theacademic program as a whole, lays emphasis on the reactor physics teaching andresearch training accorded to the students in the framework of the developed curriculum.11:35 AMMulti-Physics Nuclear Reactor Simulator for Advanced NuclearEngineering EducationAkio YamamotoNagoya University, Furo-cho, Chikusa-ku, Nagoya, JapanMulti-physics nuclear reactor simulator, which aims to utilize for advanced nuclear engineeringeducation, is being introduced to Nagoya University. The simulator consistsof the “macroscopic” physics simulator and the “microscopic” physics simulator. Theformer performs real time simulation of a whole nuclear power plant. The latter is responsibleto more detail numerical simulations based on the sophisticated and precisenumerical models, while taking into account the plant conditions obtained in the macroscopicphysics simulator. Steady-state and kinetics core analyses, fuel mechanicalanalysis, fluid dynamics analysis, and sub-channel analysis can be carried outin the microscopic physics simulator. Simulation calculations are carried out throughdedicated graphical user interface and the simulation results, i.e., spatial and temporalbehaviors of major plant parameters are graphically shown. The simulator will providea bridge between the “theories” studied with textbooks and the “physical behaviors” ofactual nuclear power plants.59

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 10:20 AM - 301 C3D - Monte Carlo Methods & DevelopmentsSession Chair: Dan Kelly (KAPL); Brian Kiedrowski (LANL)10:20 AMExplicit Temperature Treatment in Monte Carlo NeutronTracking Routines—First ResultsViitanen Tuomas and Leppänen JaakkoVTT Technical Research Centre of Finland, FI-02044 VTT, FinlandThis article discusses the preliminary implementation of the new explicit temperaturetreatment method to the development version Monte Carlo reactor physics code Serpent2 and presents the first practical results calculated using the method. The explicittemperature treatment method, as introduced in [1], is a stochastic method for takingthe effect of thermal motion into account on-the-fly in a Monte Carlo neutron transportcalculation. The method is based on explicit treatment of the motion of target nucleiat collision sites and requires cross sections at 0 K temperature only, regardless ofthe number of temperatures in the problem geometry. The method includes a novelcapability of modelling continuous temperature distributions. Test calculations are performedfor two test cases, a PWR pin-cell and a HTGR system. The resulting keff andflux spectra are compared to a reference solution calculated using Serpent 1.1.16 withDoppler-broadening rejection correction [2]. The results are in very good agreementwith the reference and also the increase in calculation time due to the new method ison acceptable level although not fully insignificant. On the basis of the current study,the explicit treatment method can be considered feasible for practical calculations.10:45 AMSpatial Homogenization of Thermal Feedback Regions inMonte Carlo Reactor CalculationsBenjamin R. Hanna, Daniel F. Gill, and David P. GriesheimerBettis Atomic Power Laboratory, Bechtel Marine Propulsion Corporation, West Mifflin, PAAn integrated thermal-hydraulic feedback module has previously been developed forthe Monte Carlo transport solver, MC21. The module incorporates a flexible input formatthat allows the user to describe heat transfer and coolant flow paths within thegeometric model at any level of spatial detail desired. The effect that the varying levelsof spatial homogenization of thermal regions has on the accuracy of the Monte Carlosimulations is examined in this study. Six thermal feedback mappings are constructedfrom the same geometric model of the Calvert Cliffs core. The spatial homogenizationof the thermal regions is varied, giving each scheme a different level of detail, andthe adequacy of the spatial homogenization is determined based on the eigenvalueproduced by each Monte Carlo calculation. The purpose of these numerical experimentsis to determine the level of detail necessarily to accurately capture the thermalfeedback effect on reactivity. Several different core models are considered: axial-flowonly, axial and lateral flow, asymmetry due to control rod insertion, and fuel heating(temperature -dependent cross sections). The thermal results generated by the MC21thermal feedback module are consistent with expectations. Based upon the numericalexperiments conducted it is concluded that the amount of spatial detail necessary toaccurately capture the feedback effect on reactivity is relatively small. Homogenizationat the assembly level for the Calvert Cliffs PWR model results in a similar power defectto that calculated with individual pin-cells modeled as explicit thermal regions.11:10 AMDevelopment of Continuous-Energy Eigenvalue SensitivityCoefficient Calculation Methods in the Shift Monte CarloCodeChristopher Perfetti and William Martin (1), Bradley Rearden and Mark Williams(2)1) University of Michigan, Department of Nuclear Engineering and Radiological Sciences, Ann Arbor, MI,USA. 2) Oak Ridge National Laboratory, Reactor and Nuclear Systems Division, Oak Ridge, TN, USAThree methods for calculating continuous-energy eigenvalue sensitivity coefficientswere developed and implemented into the Shift Monte Carlo code within the SCALEcode package. The methods were used for two small-scale test problems and wereevaluated in terms of speed, accuracy, efficiency, and memory requirements. A promisingnew method for calculating eigenvalue sensitivity coefficients, known as theCLUTCH method, was developed and produced accurate sensitivity coefficients withfigures of merit that were several orders of magnitude larger than those from existingmethods.11:35 AMSingle PIN BWR Benchmark Problem for Coupled Monte Carlo– Thermal Hydraulics AnalysisA. Ivanov, V. Sanchez (1), J.E. Hoogenboom (2)1) Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Eggenstein-Leopoldshafen. 2) Delft University of Technology, Faculty of Applied Sciences, Delft, The NetherlandsAs part of the European NURISP research project, a single pin BWR benchmark problemwas defined. The aim of this initiative is to test the coupling strategies betweenMonte Carlo and subchannel codes developed by different project participants. In thispaper the results obtained by the Delft University of Technology and Karlsruhe Instituteof Technology will be presented. The benchmark problem was simulated with the followingcoupled codes: TRIPOLISUBCHANFLOW, MCNP-FLICA, MCNP-SUBCHAN-FLOW, and KENO-SUBCHANFLOW.60

Session Chair: Bruce Bevard (ORNL)PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 10:20 AM - 301 D7B - Panel on MOXWeapons Grade MOX Fuel – A Way to a Safer FutureIn support of nonproliferation efforts, the United States is disposing of a portion of its surplus plutonium by reconstituting itinto weapons grade (WG) mixed oxide (MOX) fuel and irradiating it in commercial power reactors. Lead assemblies havebeen irradiated, and post-irradiation examination (PIE) of the assemblies is complete. The results of this work reflect anexcellent understanding of the characteristics of this fuel and how it compares to low-enriched uranium fuel and to reactorgrade (RG) MOX fuel with plutonium from recycled power reactor fuel.1.2.3.4.5.MOX PIE – Kevin McCoy – AREVA NP (discussion of MOX PIE results; tie it in to RG MOX and note differenceswith LEU if possible)Core Nuclide Report – Harold Smith - ORNL (discuss the isotopics of MOX and LEU using the SCALE/ORIGENcode report that was recently finished).Code Validation Using MOX Data – Ugur Mertyurek – ORNL (discuss what was measured and how it will beused in codes; uncertainties, what application areas these measurements can be used to validate, and somebenchmark results.Decay Heat Report – Brian Ade - ORNL (discussion of decay heat between MOX and LEU)Utility Perspective on MOX – Mick Mastilovic (TVA perspective, reactor mods etc.)61

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 10:20 AM - 301 E15A - Experimental Facilities & ExperimentsSession Chair: Tatiana Ivanova (IRSN); David Nigg (INL)10:20 AMExperimental Validation of CASMO-4E and CASMO-5M ForRadial Fission Rate Distributions in a Westinghouse SVEA-96OPTIMA2 BWR Fuel AssemblyPeter Grimm and Gregory PerretPaul Scherrer Institute, SwitzerlandMeasured and calculated radial total ssion rate distributions are compared for thethree axial sections of a Westinghouse SVEA-96 Optima2 BWR fuel assembly, comprising96, 92 and 84 fuel rods, respectively. The measurements were performed ona full-size fuel assembly in the PROTEUS zero-power experimental facility. The measuredssion rates are compared to the results of the CASMO-4E and CASMO-5M fuelassembly codes. Detailed measured geometrical data were used in the models, andeffects of the surrounding zones of the reactor were taken into account by correctionfactors derived from MCNPX calculations. The results of the calculations agree wellwith those of the experiments, with root-mean-square deviations between 1.2% and1.5% and maximum deviations of 3-4%. The quality of the predictions by CASMO-4Eand CASMO-5M is comparable.10:45 AMDecay Heat of Sodium Fast Reactor : Comparison of ExperimentalMeasurements on the PHENIX reactor with Calculationsperformed with the French DARWIN packageJ.C. Benoit, P. Bourdot, R. Eschbach, L. Boucher, V. Pascal, B. Fontaine(1), L. Martin (2), O. Serot (1)1) CEA, DEN, DER, SPRC, Cadarache, F-13108 ST Paul-Lez-Durance, France. 2) CEA, DEN, DER,Cadarache, ST Paul-Lez-Durance, France.A Decay Heat (DH) experiment on the whole core of the French Sodium-Cooled FastReactor PHENIX has been conducted in May 2008. The measurements began anhour and a half after the shutdown of the reactor and lasted twelve days. It is one ofthe experiments used for the experimental validation of the depletion code DARWINthereby confirming the excellent performance of the aforementionned code. Discrepanciesbetween measured and calculated decay heat do not exceed 8 %.11:10 AMDetermination of the Kinetic Parameters of the Caliban MetallicCore Reactor from Stochastic Neutron MeasurementsPierre Casoli, Nicolas Authier and Amaury ChapelleCommissariat à l’Energie Atomique et aux Energies Alternatives, CEA, DAM, Valduc, Is sur Tille, FranceSeveral experimental devices are operated by the Criticality and Neutron Science ResearchDepartment of the CEA Valduc Laboratory. One of these is the Caliban metalliccore reactor. The purpose of this study is to develop and perform experiments allowingto determinate some of fundamental kinetic parameters of the reactor. The promptneutron decay constant and particularly its value at criticality can be measured withreactor noise techniques such as Rossi- α and Feynman variance-to-mean methods.Subcritical, critical, and even supercritical experiments were performed. Fission chambersdetectors were put nearby the core and measurements were analyzed with theRossi-α technique. A new value of the prompt neutron decay constant at criticality wasdetermined, which allows, using the Nelson number method, new evaluations of theeffective delayed neutron fraction and the in core neutron lifetime. As an introductionof this paper, some motivations of this work are given in part 1. In part 2, principles ofthe noise measurements experiments performed at the CEA Valduc Laboratory are reminded.The Caliban reactor is described in part 3. Stochastic neutron measurementsanalysis techniques used in this study are then presented in part 4. Results of fissionchamber experiments are summarized in part 5. Part 6 is devoted to the current work,improvement of the experimental device using He 3 neutron detectors and first resultsobtained with it. Finally, conclusions and perspectives are given in part 7.11:35 AMBenchmark Experiments at Astra Facility on Definition ofSpace Distribution of 235U Fission Reaction RateA.A. Bobrov, V.F. Boyarinov, A.E. Glushkov, E.S. Glushkov, G.V. Kompaniets,N.P. Moroz, V.A. Nevinitsa, V.I. Nosov, O.N. Smirnov, P.A. Fomichenko,A.A. ZiminNational Research Centre, “Kurchatov Institute”, Moscow, RussiaResults of critical experiments performed at fiveASTRA facility configurations modelingthe high-temperature helium-cooled graphite-moderated reactors are presented.Results of experiments on definition of space distribution of 235U fission reaction rateperformed at four from these five configurations are presented more detail. Analysis ofavailable information showed that all experiments on criticality at these five configurationsare acceptable for use them as critical benchmark experiments. All experimentson definition of space distribution of 235U fission reaction rate are acceptable for usethem as physical benchmark experiments.62

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 1:30 PM - 301 A1E - Core Analysis MethodsSession Chair: W. Zwermann (GRS); Alain Santamarina (CEA)1:30 PMRoot-Cause Analysis of the Better Performance of the Coarse-Mesh Finite-Difference Method for CANDU-Type ReactorsWei ShenCandu Energy Inc., Mississauga, ON, CanadaRecent assessment results indicate that the coarse-mesh finite-difference method(FDM) gives consistently smaller percent differences in channel powers than the finemeshFDM when compared to the reference MCNP solution for CANDU-type reactors.However, there is an impression that the fine-mesh FDM should always givemore accurate results than the coarse-mesh FDM in theory. To answer the questionif the better performance of the coarse-mesh FDM for CANDU-type reactors was justa coincidence (cancelation of errors) or caused by the use of heavy water or the useof lattice-homogenized cross sections for the cluster fuel geometry in the diffusioncalculation, three benchmark problems were set up with three different fuel lattices:CANDU, HWR and PWR. These benchmark problems were then used to analyzethe root cause of the better performance of the coarse-mesh FDM for CANDU-typereactors. The analyses confirm that the better performance of the coarse-mesh FDMfor CANDU-type reactors is mainly caused by the use of lattice-homogenized crosssections for the sub-meshes of the cluster fuel geometry in the diffusion calculation.Based on the analyses, it is recommended to use 2x2 coarse-mesh FDM to analyzeCANDU-type reactors when lattice-homogenized cross sections are used in the coreanalysis.1:55 PMZEBRA: An Advanced PWR Lattice CodeLiangzhi Cao, Hongchun Wu, Youqi ZhengSchool of Nuclear Science and Technology, Xi’an Jiaotong University No. 28, Xi’an, ShannXi, P.R. ChinaThis paper presents an overview of an advanced PWR lattice code ZEBRA developedat NECP laboratory in Xi’an Jiaotong University. The multi-group cross-sectionlibrary is generated from the ENDF/B-VII library by NJOY and the 361-group SHEMstructure is employed. The resonance calculation module is developed based on subgroupmethod. The transport solver is Auto-MOC code, which is a self-developed codebased on the Method of Characteristic and the customization of AutoCAD software.The whole code is well organized in a modular software structure. Some numericalresults during the validation of the code demonstrate that this code has a good precisionand a high efficiency.2:20 PMEstimation of the Sub-Criticality of the Sodium-Cooled FastReactor Monju Using the Modified Neutron Source MultiplicationMethodG. TRUCHET (1), W.F.G. van ROOIJEN and Y. SHIMAZU (2), K. Yamaguchi(3)1) Institut National des Sciences et Techniques Nucléaires, Centre CEA de Saclay, Gif-sur-Yvette Cedex.2) Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho, Japan. 3) Japan AtomicEnergy Agency, FBR Plant Engineering Center, Tsuruga-shi, Shiraki, JAPANThe Modified Neutron Source Method (MNSM) is applied to the Monju reactor. Thisstatic method to estimate sub-criticality has already given good results on commercialPressurizedWater Reactors. The MNSM consists both in the extraction of the fundamentalmode seen by a detector to avoid the effect of higher modes near sources, andthe correction of flux distortion effects due to control rod movement. Among Monju’sparticularities that have a big influence on MNSM factors are: the presence of two californiumsources and the position of the detector which is located far from the core outsideof the reactor vessel. The importance of spontaneous fission and (α, n) reactionswhich have increased during the shutdown period of 15 years will also be discussed.The relative position of detectors and sources deeply affect the correction factors insome regions. In order to evaluate the detector count rate, an analytical propagationhas been conducted from the reactor vessel. For two subcritical states, an estimationof the reactivity has been made and compared to experimental data obtained in therestart experiments at Monju (2010).2:45 PMPseudospectral Chebyshev Representation Of Few-GroupCross Sections On Sparse GridsPavel M. Bokovand Danniëll Botes (1), Vyacheslav G. Zimin (2)1) South African Nuclear Energy Corporation (Necsa), Pretoria, South Africa. 2) Department “Automatics”,National Research Nuclear University “MEPhI”, Kashirskoe shosse, 31, RussiaThis paper presents a pseudospectral method for representing few-group homogenisedcross sections, based on hierarchical polynomial interpolation. The interpolationis performed on a multi-dimensional sparse grid built from Chebyshev nodes. Therepresentation is assembled directly from the samples using basis functions that areconstructed as tensor products of the classical one-dimensional Lagrangian interpolationfunctions. The advantage of this representation is that it combines the accuracy ofChebyshev interpolation with the efficiency of sparse grid methods. As an initial test,this interpolation method was used to construct a representation for the two-groupmacroscopic cross sections of a VVER pin cell.3:10 PMNeutronics and Depletion Methods for Parametric Studies ofFluoride-Salt-Cooled High-Temperature Reactors with SlabFuel Geometry and Multi-Batch Fuel Management SchemesAnselmo T. Cisneros (1), Dan Ilas (2)1) University of California Department of Nuclear Engineering, Berkeley, CA. 2) Oak Ridge National Laboratory,Oak Ridge, TN, USAThe Advanced High-Temperature Reactor (AHTR) is a 3,400 MWth fluoride-salt-cooledhigh-temperature reactor (FHR) that uses TRISO particle fuel compacted into slabsrather than spherical or cylindrical fuel compacts. Simplified methods are required forparametric design studies such that analyzing the entire feasible design space for anAHTR is tractable. These simplifications include fuel homogenization techniques toincrease the speed of neutron transport calculations in depletion analysis and equilibriumdepletion analysis methods to analyze systems with multi-batch fuel managementschemes. This paper presents three elements of significant novelty. First, theReactivity-Equivalent Physical Transformation (RPT) methodology usually applied insystems with coated-particle fuel in cylindrical and spherical geometries has been extendedto slab geometries. Secondly, based on this newly developed RPT method forslab geometries, a methodology that uses Monte Carlo depletion approaches wasfurther developed to search for the maximum discharge burnup in a multi-batch systemby iteratively estimating the beginning of equilibrium cycle (BOEC) compositionand sampling different discharge burnups. This Iterative Equilibrium Depletion Search(IEDS) method fully defines an equilibrium fuel cycle (keff, power, flux, and compositionevolutions) but is computationally demanding, although feasible on single-processorworkstations. Finally, an analytical method, the Non-Linear Reactivity Model, wasdeveloped by expanding the linear reactivity model to include an arbitrary number ofhigher order terms so that single-batch depletion results could be extrapolated to estimatethe maximum discharge burnup and BOEC keff in systems with multi-batch fuelmanagement schemes. Results from this method were benchmarked against equilibriumdepletion analysis results using the IEDS.63

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 1:30 PM - 301 B4E - Reactor Concepts & DesignsSession Chair: Harold Smith (ORNL)1:30 PMNeutronic Optimization in High Conversion Th-233U Fuel Assemblywith Simulated AnnealingDan Kotlyar and Eugene ShwagerausDepartment of Nuclear Engineering, Ben-Gurion University, Beer Sheva, IsraelThis paper reports on fuel design optimization of a PWR operating in a self sustainableTh-233U fuel cycle. Monte Carlo simulated annealing method was used in order toidentify the fuel assembly configuration with the most attractive breeding performance.In previous studies, it was shown that breeding may be achieved by employing heterogeneousSeed-Blanket fuel geometry. The arrangement of seed and blanket pins withinthe assemblies may be determined by varying the designed parameters based onbasic reactor physics phenomena which affect breeding. However, the amount of freeparameters may still prove to be prohibitively large in order to systematically explorethe design space for optimal solution. Therefore, the Monte Carlo annealing algorithmfor neutronic optimization is applied in order to identify the most favorable design. Theobjective of simulated annealing optimization is to find a set of design parameters,which maximizes some given performance function (such as relative period of netbreeding) under specified constraints (such as fuel cycle length). The first objective ofthe study was to demonstrate that the simulated annealing optimization algorithm willlead to the same fuel pins arrangement as was obtained in the previous studies whichused only basic physics phenomena as guidance for optimization. In the second partof this work, the simulated annealing method was used to optimize fuel pins arrangementin much larger fuel assembly, where the basic physics intuition does not yieldclearly optimal configuration. The simulated annealing method was found to be veryefficient in selecting the optimal design in both cases. In the future, this method will beused for optimization of fuel assembly design with larger number of free parametersin order to determine the most favorable trade-off between the breeding performanceand core average power density.1:55 PMNeutron Damage Reduction in a Traveling Wave ReactorAndrew G. Osborne and Mark R. DeinertDepartment of Mechanical Engineering, University of Texas at Austin, Austin, TXTraveling wave reactors are envisioned to run on depleted or natural uranium with noneed for enrichment or reprocessing, and in a manner which requires little to no operatorintervention. If feasible, this type of reactor has significant advantages over conventionalnuclear power systems. However, a practical implementation of this concept ischallenging as neutron irradiation levels many times greater than those in conventionalreactors appear to be required for a fission wave to propagate. Radiation damage tothe fuel and cladding materials presents a significant obstacle to a practical design.One possibility for reducing damage is to soften the neutron energy spectrum. Herewe show that using a uranium oxide fuel form will allow a shift in the neutron spectrumthat can result in at least a three fold decrease in dpa levels for fuel cladding andstructural steels within the reactor compared with the dpa levels expected when usinga uranium metal fuel.2:20 PMPWR Cores with Silicon Carbide CladdingJacob P. Dobisesky, David Carpenter, Edward Pilat and Mujid S. KazimiCenter for Advanced Nuclear Energy Systems, Department of Nuclear Science and Engineering, MassachusettsInstitute of Technology, Cambridge, MAThe feasibility of using silicon carbide rather than Zircaloy cladding, to reach higherpower levels and higher discharge burnups in PWRs has been evaluated. A preliminaryfuel design using fuel rods with the same dimensions as in the WestinghouseRobust Fuel Assembly but with fuel pellets having 10 vol% central void has beenadopted to mitigate the higher fuel temperatures that occur due to the lower thermalconductivity of the silicon carbide and to the persistence of the open cladpellet gapover most of the fuel life. With this modified fuel design, it is possible to achieve 18month cycles that meet present-day operating constraints on peaking factor, boronconcentration, reactivity coefficients and shutdown margin, while allowing batch averagedischarge burnups up to 80 MWD/kgU and peak rod burnups up to 100 MWD/kgU. Power uprates of 10% and possibly 20% also appear feasible. For non-upratedcores, the silicon carbide-clad fuel has a clear advantage that increases with increasingdischarge burnup. Even for comparable discharge burnups, there is a savings inenriched uranium. Control rod configuration modifications may be required to meet theshutdown margin criterion for the 20% uprate. Silicon carbide’s ability to sustain higherburnups than Zircaloy also allows the design of a licensable two year cycle with only96 fresh assemblies, avoiding the enriched uranium penalty incurred with use of largerbatch sizes due to their excessive leakage.2:45 PMOne-Group Fission Cross Sections For Plutonium and MinorActinides Inserted in Calculated Neutron Spectra of Fast ReactorCooled with Lead-208 or Leadbismuth EutecticG.L. Khorasanov, A.I. BlokhinState Scientific Center of the Russian Federation - Institute for Physics and Power, Obninsk, RussiaThe paper is dedicated to one-group fission cross sections of Pu and MA in LFRsspectra with the aim to increase these values by choosing a coolant which hardensneutron spectra. It is shown that replacement of coolant from Pb-Bi with Pb-208 in thefast reactor RBEC-M, designed in Russia, leads to increasing the core mean neutronenergy. As concerns fuel Pu isotopes, their one-group fission cross sections becomeslightly changed, while more dramatically Am-241 one-group fission cross section ischanged. Another situation occurs in the lateral blanket containing small quantities ofminor actinides. It is shown that as a result of lateral blanket mean neutron energyhardening the one-group fission cross sections of Np-237, Am-241 and Am-243 increasesup to 8- 11%. This result allows reducing the time of minor actinides burningin FRs.3:10 PMA 100MWe Advanced Sodium-Cooled Fast Reactor Core ConceptT. K. Kim, C. Grandy and R. N. HillArgonne National Laboratory, Argonne ILAn Advanced sodium-cooled Fast Reactor core concept (AFR-100) was developedtargeting a small electrical grid to be transportable to the plant site and operable fora long time without frequent refueling. The reactor power rating was strategically decidedto be 100 MWe, and the core barrel diameter was limited to 3.0 m for transportability.The design parameters were determined by relaxing the peak fast fluencelimit and bulk coolant outlet temperature to beyond irradiation experience assumingthat advanced cladding and structural materials developed under US-DOE programswould be available when the AFR-100 is deployed. With a derated power density andU-Zr binary metallic fuel, the AFR-100 can maintain criticality for 30 years without refueling.The average discharge burnup of 101 MWd/kg is comparable to conventionaldesign values, but the peak discharge fast fluence of ~6×1023 neutrons/cm2 is beyondthe current irradiation experiences with HT-9 cladding. The evaluated reactivitycoefficients provide sufficient negative feedbacks and the reactivity control systemsprovide sufficient shutdown margins. The integral reactivity parameters obtained fromquasi-static reactivity balance analysis indicate that the AFR-100 meets the sufficientconditions for acceptable asymptotic core outlet temperature following postulatedunprotected accidents. Additionally, the AFR-100 has sufficient thermal margins bygrouping the fuel assemblies into eight orifice zones.64

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 1:30 PM - 301 C3E - Monte Carlo Methods & DevelopmentsSession Chair: Dave Griesheimer (Bettis); Tom Sutton (KAPL)1:30 PMImprovements in the Mesh-Based Convergence Diagnosticsfor Monte Carlo Eigenvalue SimulationsBo Shi and Bojan PetrovicNuclear and Radiological Engineering, Georgia Institute of Technology, Atlanta, GAUsing the Monte Carlo method to compute eigenvalue problems based on the poweriteration method is widely used. However, the convergence needs accurate diagnosticsmethod in order to maintain the accuracy of the results and the efficiency of thesimulation. Recently, we used the linear regression model for convergence diagnosticsbased on the local information, the mesh-wise cycle-wise flux estimates. Our previousanalyses examined properties and demonstrated certain capability of this method.However, several aspects still need further investigation. In this paper, we use additionalsimulation examples and perform comparisons in order to better understand thecapabilities and limitations of this diagnostics method.1:55 PMAutomated Convergence Detection in Monte Carlo CriticalityCalculations Using Student’s Bridge Statistics Based on Keffor Shannon EntropyA. Jinaphanh, J. Miss, Y. Richet (1), O. Jacquet (2)1) Institute for Radiological Protection and Nuclear Safety (IRSN), Fontenay-Aux-Roses cedex - FRANCE.2) Independent ConsultantMonte Carlo (MC) criticality calculations are based on an iterative method. It requires aconverged fission source distribution before beginning tallying the effective multiplicationfactor (Keff) or other quantities of interest. However, it is pretty difficult to locateon the run, the end of the source convergence and scores may be biased by an initialtransient. This paper deals with a method that locates and suppresses the transientdue to the initialization in an output series, applied here to Keff and Shannon entropy.It relies on modeling stationary series by an order 1 auto regressive process and applyingstatistical tests based on a Student Bridge statistics. It should be noticed that theinitial transient suppression only aims at obtaining stationary output series and cannotguarantee any kind of convergence. The truncation method is applied on both Keff andShannon entropy on three test cases.2:20 PMDetermining Importance Weighting Functions for ContributonTheory Eigenvalue Sensitivity Coefficient MethodologiesChristopher Perfetti and William Martin (1), Bradley Rearden and Mark Williams(2)1) University of Michigan, Department of Nuclear Engineering and Radiological Sciences, Ann Arbor, MI,USA. 2) Oak Ridge National Laboratory, Reactor and Nuclear Systems Division, Oak Ridge, TN, USAThis study introduced two new approaches for calculating the F*(r) importance weightingfunction for Contributon and CLUTCH eigenvalue sensitivity coefficient calculations,and compared them in terms of accuracy and applicability. The necessary levelsof F*(r) mesh refinement and mesh convergence for obtaining accurate eigenvaluesensitivity coefficients were determined through two parametric studies, and the resultsof these studies suggest that a sufficiently accurate F*(r) mesh for calculatingeigenvalue sensitivity coefficients can be obtained for the Contributon and CLUTCHmethods with only a small increase in problem runtime.2:45 PMOptimization of a Coupling Scheme Between MCNP5 andSubchanflow for High Fidelity Modeling of LWR ReactorsA. Ivanov, V. Sanchez, U. Imke and K. IvanovKarlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Eggenstein-LeopoldshafenIn order to increase the accuracy and the degree of spatial resolution of core designstudies, coupled Three-Dimensional (3D) neutronics (deterministic and MonteCarlo) and 3D thermal hydraulics (CFD and sub-channel) codes are being developedworldwide. In this paper the optimization of a coupling between MCNP5 code and anin-house development thermalhydraulics code SUBCHANFLOW is presented. Variousimprovements of the coupling methodology are presented. With the help of novelinterpolation tool a consistent methodology for the preparation of thermal scatteringdata library have been developed, ensuring that inelastic scattering from bound nucleiis treated at the correct moderator temperature. Trough the utilization of a hybrid couplingwith discrete energy Monte-Carlo code KENO a methodology for acceleration ofthe coupled calculation is being demonstrated. In this approach an additional couplingbetween KENO and SUBCHANFLOW was developed, the converged results of whichare used as initial conditions for the MCNP-SUBCHANFLOW coupling. Accelerationof fission source distribution convergence, by sampling fission source distribution fromthe power distribution obtained by KENO is also demonstrated.3:10 PMUncertainty Quantification of Few Group Diffusion TheoryConstants Generated by the B1 Theoryaugumented MonteCarlo MethodHo Jin Park (1), Hyung Jin Shim, Han Gyu Joo, and Chang Hyo Kim (2)1) Korea Atomic Energy Research Institute, Daejeon, Korea. 2) Department of Nuclear Engineering,Seoul National University, Seoul, KoreaThe purpose of this paper is to quantify uncertainties of fuel pin cell or fuel assembly(FA) homogenized few group diffusion theory constants generated from the B1 theoryaugmentedMonte Carlo (MC) method. A mathematical formulation of the first kindis presented to quantify uncertainties of the few group constants in terms of the twomajor sources of the MC method; statistical and nuclear cross section and nuclidenumber density input data uncertainties. The formulation is incorporated into the SeoulNational University MC code McCARD. It is then used to compute the uncertaintiesof the burnup-dependent homogenized two group constants of a lowenriched UO2fuel pin cell and a PWR FA on the condition that nuclear cross section input data ofU-235 and U-238 from JENDL 3.3 library and nuclide number densities from the solutionto fuel depletion equations have uncertainties. The contribution of the MC inputdata uncertainties to the uncertainties of the two group constants of the two fuel systemsis separated from that of the statistical uncertainties. The utilities of uncertaintyquantifications are then discussed from the standpoints of safety analysis of existingpower reactors, development of new fuel or reactor system design, and improvementof covariance files of the evaluated nuclear data libraries.65

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 1:30 PM - 301 D7C - Fuel Cycle & Actinide ManagementSession Chair: Mick Mastilovic (TVA); Temi Taiwo (ANL)1:30 PMThe Behaviour of Transuranic Mixed Oxide Fuel in a CAN-DU-900 ReactorA. C. Morreale, M. R. Ball, D. R. Novog and J. C. LuxatDepartment of Engineering Physics, McMaster University, Ontario, CanadaThe production of transuranic actinide fuels for use in current thermal reactors providesa useful intermediary step in closing the nuclear fuel cycle. Extraction of actinidesreduces the longevity, radiation and heat loads of spent material. The burningof transuranic fuels in current reactors for a limited amount of cycles reduces the infrastructuredemand for fast reactors and provides an effective synergy that can resultin a reduction of as much as 95% of spent fuel waste while reducing the fast reactorinfrastructure needed by a factor of almost 13.5 [1]. This paper examines the featuresof actinide mixed oxide fuel, TRUMOX, in a CANDU®* nuclear reactor. The actinideconcentrations used were based on extraction from 30 year cooled spent fuel andmixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modelingwas performed using the WIMS-AECL code, super-cell calculations were analyzedin DRAGON and full core analysis was executed in the RFSP 2-group diffusion code.A time-average full core model was produced and analyzed for reactor coefficients,reactivity device worth and online fuelling impacts. The standard CANDU operationallimits were maintained throughout operations. The TRUMOX fuel design achieved aburnup of 27.36 MWd/kgHE. A full TRUMOX fuelled CANDU was shown to operatewithin acceptable limits and provided a viable intermediary step for burning actinides.The recycling, reprocessing and reuse of spent fuels produces a much more sustainableand efficient nuclear fuel cycle.1:55 PMAn Extended Conventional Fuel Cycle for the B&W mPowerSmall Modular Nuclear ReactorMichael Joseph ScarangellaThe Babcock & Wilcox Company, Lynchburg, VAThe B&W mPower reactor is a small pressurized water reactor (PWR) with an integralonce-through steam generator and a thermal output of about 500 MW; it isintended to replace aging fossil power plants of similar output. The core is composedof 69 reduced-height PWR assemblies with the familiar 17x17 fuel rod array. The Babcock& Wilcox Company (B&W) is offering a core loading and cycle management planfor a four-year cycle based on its presumed attractiveness to potential customers.This option is a once-through fuel cycle in which the entire core is discharged andreplaced after four years. In addition, a conventional fuel utilization strategy, employinga periodic partial reload and shuffle, was developed as an alternative to the four-yearonce-through fuel cycle. This study, which was performed using the Studsvik coredesign code suite, is a typical multi-cycle projection analysis of the type performed bymost fuel management organizations such as fuel vendors and utilities. In the industry,the results of such projections are used by the financial arms of these organizationsto assist in making long-term decisions. In the case of the B&W mPower reactor, thisanalysis demonstrates flexibility for customers who consider the once-through fuel cycleunacceptable from a fuel utilization standpoint. As expected, when compared to theonce-through concept, reloads of the B&W mPower reactor will achieve higher batchaverage discharge exposure, will have adequate shut-down margin, and will have arelatively flat hot excess reactivity trend at the expense of slightly increased peaking.2:20 PMProliferation Resistant Fuel for Pebble Bed Modular ReactorsYigal Ronen, Menashe Aboudy, Dror Regev and Erez GiladDepartment of Nuclear Engineering, Ben-Gurion University of the Negev, Beer-Sheva, IsraelWe show that it is possible to denature the Plutonium produced in Pebble Bed ModularReactors (PBMR) by doping the nuclear fuel with either 3050 ppm of 237Np or 2100ppm of Am vector. A correct choice of these isotopes concentration yields denaturedPlutonium with isotopic ratio 238Pu/Pu ≥ 6%, for the entire fuel burnup cycle. Thepenalty for introducing these isotopes into the nuclear fuel is a subsequent shorteningof the fuel burnup cycle, with respect to a non-doped reference fuel, by 41.2 Full PowerDays (FPDs) and 19.9 FPDs, respectively, which correspond to 4070 MWd/ton and1965 MWd/ton reduction in fuel discharge burnup.2:45 PMReactor Physics Behavior of Transuranic-Bearing Triso-ParticleFuel in a Pressurized Water ReactorMichael A. Pope, R. Sonat Sen, Abderrafi M. Ougouag, Gilles Youinou, andBrian BoerIdaho National Laboratory, Idaho Falls, ID USACalculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized WaterReactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) – only oxide fuelin tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel ofa conventional LWR after 5 years of cooling. Use of the TRISO particle fuel wouldprovide an additional barrier to fission product release in the event of cladding failure.Depletion calculations were performed to evaluate reactivity-limited burnup ofthe TRU-only FCM fuel. These calculations showed that due to relatively little spaceavailable for fuel, the achievable burnup with these pins alone is quite small. Variousreactivity parameters were also evaluated at each burnup step including moderatortemperature coefficient (MTC), Doppler, and soluble boron worth. These were comparedto reference UO2 and MOX unit cells. The TRU-only FCM fuel exhibits degradedMTC and Doppler coefficients relative to UO2 and MOX. Also, the reactivity effects ofcoolant voiding suggest that the behavior of this fuel would be similar to a MOX fuelof very high plutonium fraction, which are known to have positive void reactivity. Ingeneral, loading of TRU-only FCM fuel into an assembly without significant quantitiesof uranium presents challenges to the reactor design. However, if such FCM fuel pinsare included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivitybehavior would be dominated by the uranium pins while attractive TRU destructionperformance levels in the TRU-only FCM fuel pins is retained. From this work, it isconcluded that use of heterogeneous assemblies such as these appears feasible froma preliminary reactor physics standpoint.3:10 PMEnhanced CANDU6: Reactor and Fuel Cycle Options – NaturalUranium and BeyondM. Ovanes, P.S.W. Chan, J. Mao, N. Alderson, J.M.HopwoodCandu Energy Inc., Ontario, CanadaThe Enhanced CANDU 6® (EC6®) is the updated version of the well established CAN-DU 6 family of units incorporating improved safety characteristics designed to meet orexceed Generation III nuclear power plant expectations. The EC6 retains the excellentneutron economy and fuel cycle flexibility that are inherent in the CANDU reactordesign. The reference design is based on natural uranium fuel, but the EC6 is alsoable to utilize additional fuel options, including the use of Recovered Uranium (RU)and Thorium based fuels, without requiring major hardware upgrades to the existingcontrol and safety systems. This paper outlines the major changes in the EC6 coredesign from the existing C6 design that significantly enhance the safety characteristicsand operating efficiency of the reactor. The use of RU fuel as a transparent replacementfuel for the standard 37-el NU fuel, and several RU based advanced fuel designsthat give significant improvements in fuel burnup and inherent safety characteristicsare also discussed in the paper. In addition, the suitability of the EC6 to use MOX andrelated Pu-based fuels will also be discussed.66

Session Chair: Alain Santamarina (CEA)PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 1:30 PM - 301 E1:30 PMNuclear Data Uncertainty Analysis for the Generation IV Gas-Cooled Fast ReactorSandro Pelloni and Konstantin MikityukPaul Scherrer Institute, Villigen PSI, Switzerland11B - Nuclear Data2:20 PMResearch on Fast-Doppler-Broadening of Neutron Cross SectionsLi Songyang, Wang Kan and Yu GanglinDepartment of Engineering Physics, Tsinghua University, Beijing, ChinaFor the European 2400 MW Gas-cooled Fast Reactor (GoFastR), this paper summarizesa priori uncertainties, i.e. without any integral experiment assessment, of themain neutronic parameters which were obtained on the basis of the deterministic codesystem ERANOS (Edition 2.2-N). JEFF-3.1 cross-sections were used in conjunctionwith the newest ENDF/B-VII.0 based covariance library (COMMARA-2.0) resultingfrom a recent cooperation of the Brookhaven and Los Alamos National Laboratorieswithin the Advanced Fuel Cycle Initiative. The basis for the analysis is the original Go-FastR concept with carbide fuel pins and silicon-carbide ceramic cladding, which wasdeveloped and proposed in the first quarter of 2009 by the “French alternative energiesand Atomic Energy Commission”, CEA. The main conclusions from the current studyare that nuclear data uncertainties of neutronic parameters may still be too large forthis Generation IV reactor, especially concerning the multiplication factor, despite thefact that the new covariance library is quite complete; These uncertainties, in relativeterms, do not show the a priori expected increase with burn-up as a result of the minoractinide and fission product buildup. Indeed, they are found almost independentof the fuel depletion, since the uncertainty associated with 238U inelastic scatteringresults largely dominating. This finding clearly supports the activities of Subgroup 33of the Working Party on International Nuclear Data Evaluation Cooperation (WPEC),i.e. Methods and issues for the combined use of integral experiments and covariancedata, attempting to reduce the present unbiased uncertainties on nuclear data throughadjustments based on available experimental data.1:55 PMVerification Study of Thorium Cross Section in MVP Calculationof Thorium Based Fuel Core Using Experimental DataVu Thanh Mai, Takashi Fujii, Kazuhiro Wada and Takanori Kitada (1),Naoyuki Takaki, Akinori Yamaguchi and Hiroki Watanabe (2), HironobuUnesaki (3)1) Osaka University, Osaka, Japan. 2) Tokai University, Kanagawa, Japan. 3) Kyoto University ResearchReactor Institute, Osaka, JapanConsidering the importance of thorium data and concerning about the accuracy ofTh232 cross section library, a series of experiments of thorium critical core carried outat KUCA facility of Kyoto University Research Reactor Institute have been analyzed.The core was composed of pure thorium plates and 93% enriched uranium plates,solid polyethylene moderator with hydro to U235 ratio of 140 and Th232 to U235 ratioof 15.2. Calculations of the effective multiplication factor, control rod worth, reactivityworth of Th plates have been conducted by MVP code using JENDL-4.0 library[1]. At the experiment site, after achieving the critical state with 51 fuel rods insertedinside the reactor, the measurements of the reactivity worth of control rod and thoriumsample are carried out. By comparing with the experimental data, the calculationoverestimates the effective multiplication factor about 0.90%. Reactivity worth of thecontrol rods evaluation using MVP is acceptable with the maximum discrepancy aboutthe statistical error of the measured data. The calculated results agree to the measurementones within the difference range of 3.1% for the reactivity worth of one Th plate.From this investigation, further experiments and research on Th- 232 cross sectionlibrary need to be conducted to provide more reliable data for thorium based fuel coredesign and safety calculation.A Fast-Doppler-Broadening method is developed in this work to broaden ContinuousEnergy neutron cross-sections for Monte Carlo calculations. Gauss integration algorithmand parallel computing are implemented in this method, which is unprecedentedin the history of cross section processing. Compared to the traditional code (NJOY,SIGMA1, etc.), the new Fast-Doppler- Broadening method shows a remarkablespeedup with keeping accuracy. The purpose of using Gauss integration is to avoidcomplex derivation of traditional broadening formula and heavy load of computingcomplementary error function that slows down the Doppler broadening process. TheOpenMP environment is utilized in parallel computing which can take full advantageof modern multi-processor computers. Combination of the two can reduce processingtime of main actinides (such as 238U, 235U) to an order of magnitude of 1~2 seconds.This new method is fast enough to be applied to Online Doppler broadening. It can becombined or coupled with Monte Carlo transport code to solve temperature dependentproblems and neutronics-thermal hydraulics coupled scheme which is a big challengefor the conventional NJOY-MCNP system. Examples are shown to determine the efficiencyand relative errors compared with the NJOY results. A Godiva Benchmark isalso used in order to test the ACE libraries produced by the new method.2:45 PMSimplified Treatment of Exact Resonance Elastic ScatteringModel in Deterministic Slowing Down EquationMichitaka Ono, Kazuhiro Wada, Takanori KitadaOsaka University, 2-1, yamadaoka, Suita-shi, Osaka, JapanSimplified treatment of resonance elastic scattering model considering thermal motionof heavy nuclides and the energy dependence of the resonance cross section wasimplemented into NJOY [1]. In order to solve deterministic slowing down equation consideringthe effect of up-scattering without iterative calculations, scattering kernel forheavy nuclides is pre-calculated by the formula derived by Ouisloumen and Sanchez[2], and neutron spectrum in up-scattering term is expressed by NR approximation.To check the verification of the simplified treatment, the treatment is applied to U-238for the energy range from 4eV to 200eV. Calculated multi-group capture cross sectionof U-238 is greater than that of conventional method and the increase of the capturecross sections is remarkable as the temperature becomes high. Therefore Dopplercoefficient calculated in UO2 fuel pin is calculated more negative value than that onconventional method. The impact on Doppler coefficient is equivalent to the results ofexact treatment of resonance elastic scattering reported in previous studies [2-7].Theagreement supports the validation of the simplified treatment and therefore this treatmentis applied for other heavy nuclide to evaluate the Doppler coefficient in MOX fuel.The result shows that the impact of considering thermal agitation in resonance scatteringin Doppler coefficient comes mainly from U-238 and that of other heavy nuclidessuch as Pu239, 240 etc. is not comparable in MOX fuel.3:10 PMTENDL-2011: TALYS-based Evaluated Nuclear Data LibraryD. Rochman and A.J. KoningNuclear Research and Consultancy Group, Petten, The NetherlandsThe 4th release of the TENDL library, TENDL-2011 (TALYS-based Evaluated NuclearData Library) is described. This library consists of a complete set of nuclear reactiondata for incident neutrons, photons, protons, deuterons, tritons, helions and alpha particles,from 10−5 eV up to 200 MeV, for all isotopes from 6Li to 281Ds that are eitherstable of have a half-life longer than 1 second. All data are completely and consistentlyevaluated using a software system consisting of the TALYS-1.2 nuclear reaction code,and other programs to handle resonance data, experimental data, data from existingevaluations, and to provide the final ENDF-6 formatting. The result is a nuclear datalibrary with mutually consistent reaction information for all isotopes and a quality thatincreases with yearly updates. To produce this library, TALYS input parameters areadjusted for many nuclides so that calculated cross sections agree with experimentaldata, while for important nuclides experimental data are directly included. All informationis available on www.talys.eu and www.talys.eu/tendl-2011.67

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 3:50 PM - 301 A1F - Core Analysis MethodsSession Chair: W. Zwermann (GRS); Nan Zin Cho (KAIST);3:50 PMAnalysis of Burnup and Isotopic Compositions of BWR 9x9UO2 Fuel AssembliesM. Suzuki, T. Yamamoto, Y. Ando and T. NakajimaNuclear Energy System Safety Division, Japan Nuclear Energy Safety Organization, Tokyo, JapanIn order to extend isotopic composition data focusing on fission product nuclides,measurements are progressing using facilities of JAEA for five samples taken fromhigh burnup BWR 9x9-9 UO2 fuel assemblies. Neutronics analysis with an infiniteassembly model was applied to the preliminary measurement data using a continuousenergyMonte Carlo burnup calculation code MVP-BURN with nuclear libraries basedon JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were comparedwith those calculated using node-average irradiation histories of power and in-channelvoid fractions which were taken from the plant data. The comparison results showedthat the deviations of the calculated burnups from the measurements were -4 to 3%. Itwas confirmed that adopting the nuclear data library based on JENDL-4.0 reduced thedeviations of the calculated isotopic compositions from the measurements for 238Pu,144Nd, 145Nd, 146Nd, 148Nd, 134Cs, 154Eu, 152Sm, 154Gd, and 157Gd. On theother hand, the effect of the revision in the nuclear data library on the neutronics analysiswas not significant for major U and Pu isotopes.4:15 PMImprovements in Transport Calculations by the OptimizedMultigroup Libraries for Fast Neutron SystemsP. Mosca , C. Mounier, P. Bellier and I. ZmijarevicCommissariat à l’Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur Yvette Cedex,FranceThis paper shows how to improve the accuracy of the transport calculations using inthe APOLLO2 code the optimized multigroup libraries calculated by AEMC for fastneutron systems. These ameliorations concern the fission source calculation and theself-shielding models. The calculation of the fission source was generalized to fissionspectra including an incident neutron energy dependence. The subgroup self-shieldingmodel was updated for a mixture of resonant nuclides. Some tests on a Pu-239 spherewithout reflectors and a fast sodium cell show that the use of four fission spectraguarantees a correct representation of the fission source. The test on a Pu-239 spherewith a thick steel reflector proves that the subgroup self-shielding, accounting for themutual shielding of several resonant nuclides, allows us to improve the accuracy of theneutron transport solution in the reflector.4:40 PMMultigroup Computation of the Temperature-Dependent ResonanceScattering Model (RSM) and its ImplementationShadi Z. Ghrayeb (1), Mohamed Ouisloumen (2), Abderrafi M. Ougouag (3)and Kostadin N. Ivanov (1)1) Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, UniversityPark, PA. 2) Westinghouse Electric Company, Cranberry Township, PA, USA. 3) Idaho National Laboratory,Idaho Falls, ID, USAA multi-group formulation for the exact neutron elastic scattering kernel is developed.This formulation is intended for implementation into a lattice physics code. The correctaccounting for the crystal lattice effects influences the estimated values for the probabilityof neutron absorption and scattering, which in turn affect the estimation of corereactivity and burnup characteristics. A computer program has been written to test theformulation for various nuclides. Results of the multi-group code have been verifiedagainst the correct analytic scattering kernel. In both cases neutrons were started atvarious energies and temperatures and the corresponding scattering kernels weretallied.5:05 PMDevelopment of a Fully-Consistent Reduced Order Model toStudy Instabilities in Boiling Water ReactorsV. Dykin and C. DemaziéreChalmers University of Technology, Division of Nuclear Engineering, Department of Applied Physics,Gothenburg, SwedeA simple nonlinear Reduced Order Model to study global, regional and local instabilitiesin Boiling Water Reactors is described. The ROM consists of three submodels:neutron-kinetic, thermal-hydraulic and heat-transfer models. The neutron-kinetic modelallows representing the time evolution of the three first neutron kinetic modes: thefundamental, the first and the second azimuthal modes. The thermal-hydraulic modeldescribes four heated channels in order to correctly simulate out-of-phase behavior.The coupling between the different submodels is performed via both void and Dopplerfeedback mechanisms. After proper spatial homogenization, the governing equationsare discretized in the time-domain. Several modifications, compared to other existingROMs, have been implemented, and are reported in this paper. One novelty ofthe ROM is the inclusion of both azimuthal modes, which allows to study combinedinstabilities (in-phase and out-of-phase), as well as to investigate the correspondinginterference effects between them. The second modification concerns the precise estimationof so-called reactivity coefficients or CV;D mn - coefficients by using directcross-section data from SIMULATE-3 combined with the CORE SIM core simulator inorder to calculate eigenmodes. Furthermore, a non-uniform two-step axial power profileis introduced to simulate the separate heat production in the single and two-phaseregions, respectively. An iterative procedure was developed to calculate the solution tothe coupled neutron-kinetic/thermal-hydraulic static problem prior to solving the timedependentproblem. Besides, the possibility of taking into account the effect of localinstabilities is demonstrated in a simplified manner. The present ROM is applied to theinvestigation of an actual instability that occurred at the Swedish Forsmark-1 BWR in1996/1997. The results generated by the ROM are compared with real power plantmeasurements performed during stability tests and show a good qualitative agreement.The present study provides some insight in a deeper understanding of the physicalprinciples which drive both core-wide and local instabilities.5:30 PMDDGui, a New and Fast Way to Analyse DRAGON and DON-JON Code ResultsR. Chambon, G. MarleauInstitut de Génie Nucléaire, École Polytechnique de Montréal, Montréal, QuébecWith the largely increased performance of computer, the results from DRAGON andDONJON have increase in size and complexity. The scroll, copy and paste techniqueto get the result is not appropriate anymore. Many in-house script, software, macrohave been developed to make the data gathering easier. However, the limit of thesesolutions is their specificity and the difficulty to export them from one place to another.A general tool usable and accessible by everyone was needed. The first bricks for avery fast and intuitive way to analyse the DRAGON and DONJON results have beenput together in the graphic user interface DDGUI. Based on the extensive ROOT C++package, the possible features are numerous. For this first version of the software, wehave programmed the fundamental tools which may be the more useful on an everydaybasis: view the data structures content, draw the geometry and draw the flux orpower from a DONJON computation. The tests show how amazingly fast the user canget the informations needed for a general overview or more precise analyses. Severalother features will be implemented in the near feature.68

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 3:50 PM - 301 B2E - Deterministic Transport TheorySession Chair: Bob Grove (ORNL)3:50 PMCMFD and Coarse-Mesh DSAEdward W. Larsen and Blake W. KelleyDepartment of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MichiganUSA5:05 PMPerformance of the Block-Krylov Energy Group Solvers inJaguarA.M. Watson and R.A. KennedyKnolls Atomic Power Laboratory, Bechtel Marine Propulsion Corporation, Schenectady, New YorkThe Coarse Mesh Finite Difference (CMFD) and Diffusion Synthetic Acceleration(DSA) methods are two widely-used, independently-developed acceleration methodsfor iteratively solving deterministic particle transport simulations. In this paper we showthat these methods are related in the following way: if the standard notion of DSA asa “fine mesh” method is generalized to that of a coarse mesh method, then the linearizedform of CMFD is algebraically equivalent to a coarse mesh form of DSA. Also, wedemonstrate theoretically (via Fourier analysis) and computationally that CMFD andcoarse mesh DSA have nearly the same convergence properties.A new method of coupling the inner and outer iterations for deterministic transportproblems is proposed. This method is termed the Multigroup Energy Blocking Method(MEBM) and has been implemented in the deterministic transport solver Jaguar, whichis currently under development at KAPL. The method is derived for both fixed-sourceand eigenvalue problems. The method is then applied to a PWR pin cell model, both infixed-source mode and eigenvalue mode. The results show that the MEBM improvesthe convergence of both types of problems when applied to the thermal (upscattering)groups.4:15 PMA Parallel in Energy Krylov Solution Algorithm for the VariationalNodal MethodYunzhao Li (1), E. E. Lewis (2), Micheal A. Smith (3)1) School of Nuclear Science and Technology, Xi’an Jiaotong University, Shaanxi, China. 2) Departmentof Mechanical Engineering, Northwestern University, Evanston, IL, USA. 3) Nuclear Engineering Division,Argonne National Laboratory, Argonne, IL, USATo reduce the computational expense caused by the large number of energy groupsin thermal and fast reactor calculations, we researched using parallelization in energy(i.e. group) of the multigroup response matrix equations derived from the VariationalNodal Method. This study focuses on solving steady-state eigenvalue problems indiffusion theory that are typically needed for depletion and reactivity coefficient calculations.To examine serial and parallel performance, GMRES, a Krylov algorithm,is applied to the multi-group system of equations at each fission source iteration. Wealso study the impact of using Wielandt acceleration on the eigenvalue/eigenvectorproblem. Numerical results based on a 72 group thermal and a 216 group fast reactorcore problems are obtained for both serial and parallel calculations. The results showreduced CPU times when using GMRES and Wielandt acceleration in both serial andparallel calculations.5:30 PMPost Irradiation Experiment Analysis using the APOLLO2 deterministictool. Validation of JEFF-3.1.1 thermal and epithermalactinides neutron induced cross sections through ME-LUSINE experimentsDavid Bernard and Olivier FabbrisCEA, DEN, DER, SPRC, Laboratoire d’Etudes de Physique, Saint Paul Lez Durance, FranceTwo different experiments performed in the 8MWth MELUSINE experimental powerpool reactor aimed at analyzing 1GWd/t spent fuel pellets doped with several actinides.The goal was to measure the averaged neutron induced capture cross section in twovery different neutron spectra (a PWR-like and an under-moderated one). This papersummarizes the combined deterministic APOLLO2-stochastic TRIPOLI4 analysis usingthe JEFF-3.1.1 European nuclear data library. A very good agreement is observedfor most of neutron induced capture cross section of actinides and a clear underestimationfor the 241Am(n,g) as an accurate validation of its associated isomeric ratio areemphasized. Finally, a possible huge resonant fluctuation (factor of 2.7 regarding tothe l=0 resonance total orbital momenta) is suggested for isomeric ratio.4:40 PMA Non-Linear Discontinuous Petrov-Galerkin Method for RemovingOscillations in the Solution of the Time-DependentTransport EquationS. R. Merton and R. P. Smedley-Stevenson (1), C. C. Pain (2)1) Computational Physics Group, AWE Aldermaston,, Berkshire, United Kingdom. 2) Department of EarthScience and Engineering, Imperial College London, London, United KingdonThis paper describes a Non-Linear Discontinuous Petrov-Galerkin method and its applicationto the one-speed Boltzmann Transport Equation (BTE) for space-time problems.The purpose of the method is to remove unwanted oscillations in the transportsolution which occur in the vicinity of sharp flux gradients, while improving computationalefficiency and numerical accuracy. This is achieved by applying artificial dissipationin the solution gradient direction, internal to an element using a novel finiteelement (FE) Riemann approach. The added dissipation is calculated at each nodeof the finite element mesh based on local behaviour of the transport solution on boththe spatial and temporal axes of the problem. Thus a different dissipation is used indifferent elements. The magnitude of dissipation that is used is obtained from a gradient-informedscaling of the advection velocities in the stabilisation term. This makesthe method in its most general form non-linear. The method is implemented within avery general finite element Riemann framework. This makes it completely independentof choice of angular basis function allowing one to use different descriptions of theangular variation. Results show the non-linear scheme performs consistently well indemanding time-dependent multi-dimensional neutron transport problems.69

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 3:50 PM - 301 C6A - Reactor Operation & SafetySession Chair: Dumitru Serghiuta (CNSC)3:50 PMReactor Physics Studies for Assessment of Tramp UraniumMethodsP. Grimm, A. Vasiliev, W. Wieselquist, H. Ferroukhi (1), G. Ledergerber (2)1) Paul Scherrer Institut, Villigen PSI, Switzerland. 2) Kernkraftwerk Leibstadt AG, Leibstadt, Switzerland5:05 PMDecay-Ratio Calculation in the Frequency Domain with theLAPUR Code Using 1D-KineticsJ.L. Munoz-Cobo, A. Escrivá, C. García, C. Berna (1) J. Melara (2)1) Instituto de Ingeniería Energética, Universitat Politécnica de Valencia, Camino de Vera s/n Valencia,Spain. 2) IBERDROLA Ingeniería y Construcción, Madrid, SpainThis paper presents calculation studies towards validation of a methodology for estimationsof the tramp uranium mass from water chemistry measurements. Particularemphasis is given to verify, from a reactor physics point of view, the justification basisfor the so-called “Pu-based model” versus the “U-based model” as a key assumptionfor the methodology. The computational studies are carried out for a typical BWR fuelassembly with CASMO-5M and MCNPX. By approximating the evolution of fissile nuclidesand the fraction of 235U fissions to total fissions in different zones of a fuel rod,including tramp uranium on the clad surface, it is found that Pu gives the dominantcontribution to fissions for tramp uranium after an irradiation on the outer clad surfaceof at least one cycle in a BWR. Thus, the use of the so-called Pu model for the determinationof the tramp uranium mass (this means in particular using the yields for 239Pufission) appears justified in the cases considered. On that basis, replacing the older Umodel by a Pu model is recommended.This paper deals with the problem of computing the Decay Ratio in the frequencydomain codes as the LAPUR code. First, it is explained how to calculate the feedbackreactivity in the frequency domain using slab-geometry i.e. 1D kinetics, also we showhow to perform the coupling of the 1D kinetics with the thermal-hydraulic part of theLAPUR code in order to obtain the reactivity feedback coefficients for the differentchannels. In addition, we show how to obtain the reactivity variation in the complexdomain by solving the eigenvalue equation in the frequency domain and we comparethis result with the reactivity variation obtained in first order perturbation theory usingthe 1D neutron fluxes of the base case. Because LAPUR works in the linear regime, itis assumed that in general the perturbations are small. There is also a section devotedto the reactivity weighting factors used to couple the reactivity contribution from thedifferent channels to the reactivity of the entire reactor core in point kinetics and 1Dkinetics. Finally we analyze the effects of the different approaches on the DR value.4:15 PMComputer Code for Space-Time Diagnostics of Nuclear SafetyParametersD.A. Solovyev, A.A. Semenov, F.V. Gruzdov, A.A. Druzhaev, N.V. Shchukin,S.G. Dolgenko, I.V. Solovyeva, E.A. OvchinnikovaNational Research Nuclear University “MEPhI”, Moscow, RussiaThe computer code ECRAN 3D (Experimental & Calculation Reactor Analysis) isdesigned for continuous monitoring and diagnostics of reactor cores and databasesfor RBMK-1000 on the basis of analytical methods for the interrelation parametersof nuclear safety. The code algorithms are based on the analysis of deviations betweenthe physically obtained figures and the results of neutron-physical and thermalhydrauliccalculations. Discrepancies between the measured and calculated signalsare equivalent to obtaining inadequacy between performance of the physical deviceand its simulator. The diagnostics system can solve the following problems: identificationof facts and time for inconsistent results, localization of failures, identification andquantification of the causes for inconsistencies. These problems can be effectivelysolved only when the computer code is working in a real-time mode. This leads toincreasing requirements for a higher code performance. As false operations can leadto significant economic losses, the diagnostics system must be based on the certifiedsoftware tools. POLARIS, version 4.2.1 is used for the neutron-physical calculation inthe computer code ECRAN 3D.5:05 PMSteam Turbine: Alternative Emergency Drive for the SecureRemoval of Residual Heat from the Core of Light Water Reactorsin Ultimate Emergency SituationRubens Souza dos Santos (1,2)1) Instituto de Engenharia Nuclear (CNEN/IEN), Rio de Janeiro, Brazil. 2) Instituto Nacional de Ciência eTecnologia de Reatores Nucleares Inovadores / CNPqIn 2011 the nuclear power generation has suffered an extreme probation. That couldbe the meaning of what happened in Fukushima Nuclear Power Plants. In thoseplants, an earthquake of 8.9 on the Richter scale was recorded. The quake intensitywas above the trip point of shutting down the plants. Since heat still continued tobe generated, the procedure to cooling the reactor was started. One hour after theearthquake, a tsunami rocked the Fukushima shore, degrading all cooling system ofplants. Since the earthquake time, the plant had lost external electricity, impacting thepumping working, drive by electric engine. When operable, the BWR plants respondedthe management of steam. However, the lack of electricity had degraded the plant maneuvers.In this paper we have presented a scheme to use the steam as an alternativedrive to maintain operable the cooling system of nuclear power plant. This schemeadds more reliability and robustness to the cooling systems. Additionally, we purposeda solution to the cooling in case of lacking water for the condenser system. In our approach,steam driven turbines substitute electric engines in the ultimate emergencycooling system.70

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 3:50 PM - 301 D14A - Reactor Transient & Safety AnalysisSession Chair: Tom Sutton (KAPL); Sara Bortot (KTH)3:50 PMDevelopment of Refined MCNPX-PARET Multi-Channel Modelfor Transient Analysis in Research ReactorsS. Kalcheva and E. Koonen (1), A. P. Olson (2)1) SCK•CEN, BR2 Reactor Department, Mol – Belgium, 2) RERTR Program, Argonne National Laboratory,Argonne, IL – USAReactivity insertion transients are often analyzed (RELAP, PARET) using a two-channelmodel, representing the hot assembly with specified power distribution and anaverage assembly representing the remainder of the core. For the analysis of protectedby the reactor safety system transients and zero reactivity feedback coefficientsthis approximation proves to give adequate results. However, a more refined multichannelmodel representing the various assemblies, coupled through the reactivityfeedback effects to the whole reactor core is needed for the analysis of unprotectedtransients with excluded over power and period trips. In the present paper a detailedmulti-channel PARET model has been developed which describes the reactor core indifferent clusters representing typical BR2 fuel assemblies. The distribution of powerand reactivity feedback in each cluster of the reactor core is obtained from a bestestimateMCNPX calculation using the whole core geometry model of the BR2 reactor.The sensitivity of the reactor response to power, temperature and energy distributionsis studied for protected and unprotected reactivity insertion transients, with zero andnon-zero reactivity feedback coefficients. The detailed multi-channel model is comparedvs. simplified fewer-channel models. The sensitivities of transient characteristicsderived from the different models are tested on a few reactivity insertion transients withreactivity feedback from coolant temperature and density change.4:15 PMTrace/Parcs Modelling of RIPS Trip Transients for LUNGMENABWRChia-Ying Chang (1), Hao-Tzu Lin and Jong-Rong Wang (2), ChunkuanShih (1)1)Institute of Nuclear Engineering and Science, National Tsing-Hua University, Hsinchu, Taiwan, 2) Instituteof Nuclear Energy Research, Taiwan, Taoyuan County, TaiwanThe objectives of this study are to examine the performances of the steady-state resultscalculated by the Lungmen TRACE/PARCS model compared to SIMULATE-3code, as well as to use the analytical results of the final safety analysis report (FSAR)to benchmark the Lungmen TRACE/PARCS model. In this study, three power generationmethods in TRACE were utilized to analyze the three reactor internal pumps(RIPs) trip transient for the purpose of validating the TRACE/PARCS model. In general,the comparisons show that the transient responses of key system parametersagree well with the FSAR results, including core power, core inlet flow, reactivity, etc.Further studies will be performed in the future using Lungmen TRACE/PARCS model.After the commercial operation of Lungmen nuclear power plant, TRACE/PARCSmodel will be verified.4:40 PMComparison of Point Kinetics, Improved Quasistatic andTheta Method as Space-Time Kinetics Solvers in DONJON-3SimulationsR. Chambon, G. MarleauInstitut de Génie Nucléaire, École Polytechnique de Montréal, Montréal, Québec, CANADATo ensure the safety of nuclear reactors, we have to simulate accurately their normaloperation and also accident cases. To perform transient calculations, coupled neutronicand thermohydaulic codes are used. This article compares three neutronic solvers.The first one is the point kinetic approach where the flux shape is constant duringall the transient. For the second method (the improved quasistatic method), the fluxshape is constant but only during small time steps. Finally, we used the theta approachwhere both flux and precursors distributions vary with time and space. Transients ofLost Of Coolant Accident in CANDU-6 reactors have been simulated with DONJONand the outputs of a thermalhydraulic system code. Results show that the point kineticsis inappropriate for transient with large distortion of the flux shape. Improvedquasistatic and theta methods give relatively similar results. However, the improvedquasistatic approach is less stable and a little bit more sensitive on time-step andspatial discretization than the theta method is.5:05 PM3D Analysis of the Reactivity Insertion Accident In VVER-1000Abdullayev A.M., Zhukov A.I., and Slyeptsov S.M.NSC “Kharkov Institute for Physics and Technology”, Kharkov, UkraineFuel parameters such as peak enthalpy and temperature during rod ejection accidentare calculated. The calculations are performed by 3D neutron kinetics code NESTLEand 3D thermal-hydraulic code VIPRE-W. Both hot zero power and hot full power caseswere studied for an equilibrium cycle with Westinghouse hex fuel in VVER-1000. Itis shown that the use of 3D methodology can significantly increase safety margins forcurrent criteria and met future criteria.5:30 PMCPR Methodology with New Steady-State Criterion and MoreAccurate Statistical Treatment of Channel BOWS. Baumgartner (1), R. Bieli (2), U.C. Bergmann (3)1) Axpo AG, Baden, Switzerland. 2) Kernkraftwerk Leibstadt AG, Leibstadt, Switzerland. 3) WestinghouseElectric Sweden AB, Västerås, SwedenAn overview is given of existing CPR design criteria and the methods used in BWRreload analysis to evaluate the impact of channel bow on CPR margins. Potentialweaknesses in today’s methodologies are discussed. Westinghouse in collaborationwith KKL and Axpo – operator and owner of the Leibstadt NPP – has developed anoptimized CPR methodology based on a new criterion to protect against dryout duringnormal operation and with a more rigorous treatment of channel bow. The newsteady-state criterion is expressed in terms of an upper limit of 0.01 for the dryoutfailure probability per year. This is considered a meaningful and appropriate criterionthat can be directly related to the probabilistic criteria set-up for the analyses of AnticipatedOperation Occurrences (AOOs) and accidents. In the Monte Carlo approach astatistical modeling of channel bow and an accurate evaluation of CPR response functionsallow the associated CPR penalties to be included directly in the plant SLMCPRand OLMCPR in a best-estimate manner. In this way, the treatment of channel bow isequivalent to all other uncertainties affecting CPR. Emphasis is put on quantifying thestatistical distribution of channel bow throughout the core using measurement data.The optimized CPR methodology has been implemented in the Westinghouse MonteCarlo code, McSLAP. The methodology improves the quality of dryout safety assessmentsby supplying more valuable information and better control of conservatisms inestablishing operational limits for CPR. The methodology is demonstrated with applicationexamples from the introduction at KKL.71

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 - 3:50 PM - 301 E8D - Advanced Modeling & Simulation in Reactor PhysicsSession Chair: Akio Yamamoto (Nagoya U); K. Ivanov (PSU)3:50 PMIRSN Working Program Status on Tools for Evaluation of SFRCores Static Neutronics Safety ParametersE. Ivanov, V. Tiberi, F. Ecrabet, Y. Chegrani, E. Canuti, D. Bisogni, A. SargeniAnd F. BernardInstitut de Radioprotection et de Sûreté Nucléaire (IRSN), Fontenay-aux-rosesAs technical support of the French Nuclear Safety Authority, IRSN will be in chargeof safety assessment of any future project of Sodium Fast Reactor (SFR) that couldbe built in France. One of the main safety topics will deal with reactivity control. Sincethe design and safety assessment of the last two SFR plants in France (Phénix andSuperphénix, more than thirty years ago), methods, codes and safety objectives haveevolved. That is why a working program on core neutronic simulations has beenlaunched in order to be able to evaluate accuracy of future core characteristics computations.The first step consists in getting experienced with the ERANOS well-knowndeterministic code used in the past for Phénix and Superphénix. Then Monte-Carlocodes have been tested to help in the interpretation of ERANOS results and to definewhat place this kind of codes can have in a new SFR safety demonstration. This experienceis based on open benchmark computations. Different cases are chosen to covera wide range of configurations. The paper shows, as an example, criticality resultsobtained with ERANOS, SCALE and MORET, and the first conclusions based on theseresults. In the future, this work will be extended to other safety parameters such assodium void and Doppler effects, kinetic parameters or flux distributions.4:15 PMComputation Of Neutron Fluxes in Clusters of Fuel Pins Arrangedin Hexagonal Assemblies (2D And 3D)Hem Prabha and Guy MarleauInstitut de Génie Nucléaire, École Polytechnique de Montréal, Montréal, QuébecFor computations of fluxes, we have used Carvik’s method of collision probabilities.This method requires tracking algorithms. An algorithm to compute tracks (in 2D and3D) has been developed for seven hexagonal geometries with cluster of fuel pins. Thishas been implemented in the NXT module of the code DRAGON. The flux distributionin cluster of pins has been computed by using this code. For testing the results, theyare compared when possible with the EXCELT module of the code DRAGON. Tracksare plotted in the NXT module by using MATLAB, these plots are also presented here.Results are presented with increasing number of lines to show the convergence ofthese results. We have numerically computed volumes, surface areas and the percentageerrors in these computations. These results show that 2D results convergefaster than 3D results. The accuracy on the computation of fluxes up to second decimalis achieved with fewer lines.4:40 PMThe Improvement of the Method of Equivalent Cross Sectionin HTRGuo Jiong, Li FuInstitute of Nuclear and New Energy Technology, Tsinghua University, Beijing, ChinaThe Method of Equivalence Cross-Sections (MECS) is a combined transport-diffusionmethod. By appropriately adjusting the diffusion coefficient of homogenized absorberregion, the diffusion theory could yield satisfactory results for the full core model withstrong neutron absorber material, for example the control rod in High temperature gascooled reactor (HTR). Original implementation of MECS based on 1-D cell transportmodel has some limitation on accuracy and applicability, a new implementation ofMECS based on 2-D transport model are proposed and tested in this paper. This improvementcan extend the MECS to the calculation of twin small absorber ball systemwhich have a non-circular boring in graphite reflector and different radial position. Aleast-square algorithm for the calculation of equivalent diffusion coefficient is adopted,and special treatment for diffusion coefficient for higher energy group is proposed inthe case that absorber is absent. Numerical results to adopt MECS into control rodcalculation in HTR are encouraging. However, there are some problems left.5:05 PMConstruction of Accuracy-Preserving Surrogate for the EigenvalueRadiation Diffusion and/or Transport ProblemCongjian Wang and Hany S. Abdel-KhalikDepartment of Nuclear Engineering, North Caroline State University, Raleigh, NCThe construction of surrogate models for high fidelity models is now considered animportant objective in support of all engineering activities which require repeated executionof the simulation, such as verification studies, validation exercises, and uncertaintyquantification. The surrogate must be computationally inexpensive to allow itsrepeated execution, and must be computationally accurate in order for its predictionsto be credible. This manuscript introduces a new surrogate construction approach thatreduces the dimensionality of the state solution via a range-finding algorithm from linearalgebra. It then employs a proper orthogonal decompositionlike approach to solvefor the reduced state. The algorithm provides an upper bound on the error resultingfrom the reduction. Different from the state-of-the-art, the new approach allows theuser to define the desired accuracy a priori which controls the maximum allowablereduction. We demonstrate the utility of this approach using an eigenvalue radiationdiffusion model, where the accuracy is selected to match machine precision. Resultsindicate that significant reduction is possible for typical reactor assembly models,which are currently considered expensive given the need to employ very fine meshmany group calculations to ensure the highest possible fidelity for the downstreamcore calculations. Given the potential for significant reduction in the computationalcost, we believe it is possible to rethink the manner in which homogenization theory iscurrently employed in reactor design calculations.5:30 PM3D Neutronic/Thermal-Hydraulic Coupled Analysis of MYR-RHAMiriam Vazquez, Francisco Martin-FuertesCIEMAT, Madrid, SpainThe current tendency in multiphysics calculations applied to reactor physics is theuse of already validated computer codes, coupled by means of an iterative approach.In this paper such an approach is explained concerning neutronics and thermal-hydraulicscoupled analysis with MCNPX and COBRA-IV codes using a driver programand file exchange between codes. MCNPX provides the neutronic analysis of heterogeneousnuclear systems, both in critical and subcritical states, while COBRA-IVis a subchannel code that can be used for rod bundles or core thermal-hydraulicsanalysis. In our model, the MCNP temperature dependence of nuclear data is handledvia pseudo-material approach, mixing pre-generated cross section data set to obtainthe material with the desired cross section temperature. On the other hand, COBRA-IV has been updated to allow for the simulation of liquid metal cooled reactors. Thecoupled computational tool can be applied to any geometry and coolant, as it is thecase of single fuel assembly, at pin-by-pin level, or full core simulation with the averagepin of each fuel-assembly. The coupling tool has been applied to the critical core layoutof the SCK-CEN MYRRHA concept, an experimental LBE cooled fast reactor presentlyin engineering design stage.72

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationWednesday April 18, 2012 -6:30 PM - Ballroom A-DWednesday BanquetDr. H. Lee Dodds (Honorary Chair)IBM Professor of Engineering and Department Head, EmeritusUniversity of Tennessee Nuclear Engineering DepartmentModerator: “PHYSOR 2012 Student Best Paper Awards”Introduction by Ivan MaldonadoDr. H. L. (Lee) Dodds joined UTNE in 1976 after working for the DuPont Company at the Savannah RiverLaboratory for six years. He began serving as UTNE Department Head in early 1997 and retired in Decemberof 2011. He also previously worked at the Oak Ridge National Laboratory and the National Aeronauticsand Space Administration. Dr. Dodds earned B.S., M.S., and Ph.D. degrees in nuclear engineering at UT.He has served as a consultant for the U.S. Department of Energy and several American, Canadian, and Dutch research institutes andcompanies. He currently serves on the External Advisory Boards for the nuclear engineering programs at Ohio State University and VirginiaCommonwealth University. He is also a member of the Accreditation Board of the National Academy for Nuclear Training, a memberof the National Board of Directors of the American Nuclear Society (ANS), and a past member of the National Board of Directors of theNuclear Energy Institute (NEI). Dr. Dodds has received many awards including the ANS Arthur Holly Compton National Teaching Award.He is a Licensed Professional Engineer in Tennessee and a Fellow of ANS.Dr. Paul J. Turinsky (Honorary Chair)Professor of Nuclear EngineeringNorth Carolina State UniversityPresentation: “Overview of the CASL Project”Introduction by Ivan MaldonadoDr. Paul Turinsky is a Professor of Nuclear Engineering at North Carolina State University and serves asChief Scientist for the Consortium for Advanced Simulation of Light Water Reactors (CASL), a Departmentof Energy Innovation Hub. His research interests include computational reactor physics, nuclear fuel management, uncertainty quantification and data assimilation. At NC State he also serves as faculty coordinatorof the Interdisciplinary Graduate Program in Computational Engineering and Science. He is a Fellow of the American Nuclear Societyand recipient of several awards, including the ANS’s Arthur Holly Compton Award in Education and Eugene P. Wigner Reactor PhysicistAward, and the Department of Energy’s E. O. Lawrence Award.73

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 8:00 AM - 301 A6B - Reactor Operation & SafetySession Chair: K. Kozier (AECL)8:00 AMEvaluation of Accuracy of Calculations of VVER-1000 CoreStates with Incomplete Covering of Fuel by the AbsorberA.V.Tikhomirov, G.L.PonomarenkoOKB “GIDROPRESS”, Podolsk, RussiaAn additional verification of bundled software (BS) SAPFIR_95&RC [1] and code KOR-SAR/GP [2] was performed. Both software products were developed in A.P. AlexandrovNITI and certified by ROSTEKHNADZOR of RF for numeric simulation of stationary,transitional and emergency conditions of VVER reactors. A benchmark model forneutronics calculations was created within the limits of this work. The cold subcriticalstate of VVER - 1000 reactor stationary fuelling was simulated on the basis of FA withan increased height of the fuel column (TVS-2М) considering detailed presentation ofradial and front neutron reflectors. A case of passing of pure condensate slug throughthe core in initially deep subcritical state during start of the first RCP set after refuelingwas considered as an examined condition of reactor operation. A relatively small sizeof the slug, its spatial position near the reflectors (lower and lateral), as well as failureof the inserted control rods of the control and protection system (CPS CR) to reach thelower limit of the fuel column stipulate for methodical complexity of a correct calculationof the neutron multiplication constant (Кeff) using engineering codes. Code RCwas used as a test program in the process of reactor calculated 3-D modeling. CodeMCNP5 [3] was used as the precision program, which solves the equation of neutronstransfer by Monte-Carlo method and which was developed in the US (Los-Alamos).As a result of comparative calculations dependency of Кeff on two parameters wasevaluated – boron acid concentration (Cb) and CPS CR position. Reactivity effect wasevaluated, which is implemented as a result of failure of all CPS control rods to reachthe lower fuel limit calculated using the engineering codes mentioned above.8:25 AMImprovement of the Thermal Margins in the Swedish Ringhals-3PWR by Introducing New Fuel Assemblies with ThoriumCheuk Wah Lau and Christophe Demazière (1), Henrik Nylén and UrbanSandberg (2)1) Department of Applied Physics, Division of Nuclear Engineering, Chalmers University of Technology,Gothenburg, Sweden. 2) Ringhals AB, Väröbacka, SwedenThorium is a fertile material and most of the past research has focused on breedingthorium to fissile material. In this paper, the focus is on using thorium to improve thethermal margins by homogeneously distributing thorium in the fuel pellets. A proposeduranium-thorium-based fuel assembly is simulated for the Swedish Ringhals-3 PWRcore in a realistic demonstration. All the key safety parameters, such as isothermaltemperature coefficient of reactivity, Doppler temperature of reactivity, boron worth,shutdown margins and fraction of delayed neutrons are studied in this paper, andare within safety limits for the new core design using the uraniumthorium- based fuelassemblies. The calculations were performed by the two-dimensional transport codeCASMO-4E and the two group steady-state three dimensional nodal code SIMU-LATE-3 from Studsvik Scandpower. The results showed that the uranium-thoriumbasedfuel assembly improves the thermal margins, both in the pin peak power andthe local power (Fq). The improved thermal margins would allow more flexible coredesigns with less neutron leakage or could be used in power uprates to offer efficientsafety margins.8:50 AMA Reactor Core On-Line Monitoring Program - COMPWang Changhui (1,2), Wu Hongchun, Cao Liangzhi (2)1) State Nuclear Power Software Development Center, Beijing, P.R. China. 2) School of Nuclear Scienceand Technology, Xi’an Jiaotong UniversityA program named COMP is developed for on-line monitoring PWRs’ in-core powerdistribution in this paper. Harmonics expansion method is used in COMP. The Unit 1reactor of DayaBay Nuclear Power Plant (DayaBay NPP) in China is considered forverification. The numerical results show that the maximum relative error between measurementand reconstruction results from COMP is less than 5%, and the computingtime is short, indicating that COMP is capable for on-line monitoring PWRs.9:15 AMA Simplified Spatial Model for BWR StabilityYonatan Berman (1), Yoav Lederer (2), Ehud Meron (3)1) Department of Physics, Ben-Gurion University of the Negev and Nuclear Research Center-Negev,Beer-Sheva, Israel. 2) Department of Physics, Nuclear Research Center-Negev, Beer-Sheva, Israel. 3)Department of Solar Energy and Environmental Physics and Department of Physics, Ben-Gurion Universityof the Negev, Beer-Sheva, IsraelA spatial reduced order model for the study of BWR stability, based on the phenomenologicalmodel of March-Leuba et al., is presented. As one dimensional spatial dependenceof the neutron flux, fuel temperature and void fraction is introduced, it ispossible to describe both global and regional oscillations of the reactor power. Bothlinear stability analysis and numerical analysis were applied in order to describe theparameters which govern the model stability. The results were found qualitatively similarto past results. Doppler reactivity feedback was found essential for the explanationof the different regions of the flow-power stability map.9:40 AMSurveillance and Diagnostics of The Beam Mode Vibrationsof the Ringhals PWRsCristina Montalvo Martín (1), Imre Pázsit (2), Henrik Nylén (3)Spain. 2) Department of Nuclear Engineering, Chalmers University of Technology, Göteborg, Sweden. 3)Vattenfall Ringhals AB, Väröbacka, SwedenSurveillance of core barrel vibrations has been performed in the Swedish RinghalsPWRs for several years. This surveillance is focused mainly on the pendular motion ofthe core barrel, which is known as the beam mode. The monitoring of the beam modehas suggested that its amplitude increases along the cycle and decreases after refuelling.In the last 5 years several measurements have been taken in order to understandthis behaviour. Besides, a non-linear fitting procedure has been implemented in orderto better distinguish the different components of vibration. By using this fitting procedure,two modes of vibration have been identified in the frequency range of the beammode. Several results coming from the trend analysis performed during these yearsindicate that one of the modes is due to the core barrel motion itself and the other isdue to the individual flow induced vibrations of the fuel elements. In this work, the latestresults of this monitoring are presented.74

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 8:00 AM - 301 B15B - Experimental Facilities & ExperimentsSession Chair: Frederik Reitsma(Calvera consultants); Alain Santamarina (CEA)8:00 AMPreserving Physics Knowledge at the Fast Flux Test FacilityDavid Wootan and Ronald Omberg (1), Bruce J. Makenas (2), Deborah L.Nielsen and Joseph V. Nelson (3), David L. Polzin (4)1) Pacific Northwest National Laboratory, Richland, Washington. 2) Ares Corporation, Richland, Washington.3) Indian Eyes, LLC, Pasco, Washington. 4) CH2MHill Plateau Remediation Company, Richland,WashingtonOne of the goals of the Department of Energy’s Office of Nuclear Energy, initiatedunder the Fuel Cycle Research and Development Program (FCRD) and continuedunder the Advanced Reactor Concepts Program (ARC) is to preserve the knowledgethat has been gained in the United States on Liquid Metal Reactors (LMRs) that couldsupport the development of an environmentally and economically sound nuclear fuelcycle. The Fast Flux Test Facility (FFTF) is the most recent LMR to operate in theUnited States, from 1982 to 1992, and was designed as a fully instrumented test reactorwith on-line, real time test control and performance monitoring of components andtests installed in the reactor. The 10 years of operation of the FFTF provided a veryuseful framework for testing the advances in LMR safety technology based on passivesafety features that may be of increased importance to new designs after the eventsat Fukushima. Knowledge preservation at the FFTF is focused on the areas of design,construction, and startup of the reactor, as well as on preserving information obtainedfrom 10 years of successful operating history and extensive irradiation testing of fuelsand materials. In order to ensure protection of information at risk, the program to datehas sequestered reports, files, tapes, and drawings to allow for secure retrieval. Adisciplined and orderly approach has been developed to respond to client’s requestsfor documents and data in order to minimize the search effort and ensure that futurerequests for this information can be readily accommodated.8:50 AMThe Features of Neutronic Calculations for Fast Reactorswith Hybrid Cores on the Basis of BFS-62-3A Critical AssemblyExperimentsE. F. Mitenkova and N. V. Novikov (1), A. I. Blokhin (2)1) Nuclear Safety Institute of Russian Academy of Sciences, Moscow, Russia. 2) State Scientific Centerof Russian Federation - Institute of Physics and Power Engineering named after A.I. Leypunsky, Obninsk,RussiaThe different (U-Pu) fuel compositions are considered for next generation of sodiumfast breeder reactors. The considerable discrepancies in axial and radial neutron spectrafor hybrid reactor systems compared to the cores with UO2 fuel cause increasinguncertainty of generating the group nuclear constants in those reactor systems. Thecalculation results of BFS-62-3A critical assembly which is considered as full-scalemodel of BN-600 hybrid core with steel reflector specify quite different spectra in localareas. For those systems the MCNP 5 calculations demonstrate significant sensitivityof effective multiplication factor Keff and spectral indices to nuclear data libraries. For235U, 238U, 239Pu the results of calculated radial fission rate distributions against thereconstructed ones are analyzed. Comparative analysis of spectral indices, neutronspectra and radial fission rate distributions are performed using the different versionsof ENDF/B, JENDL-3.3, JENDL-4, JEFF-3.1.1 libraries and BROND-3 for Fe, Cr isotopes.For analyzing the fission rate sensitivity to the plutonium presence in the fuel239Pu is substituted for 235U (enrichment 90%) in the FA areas containing the plutonium.For 235U, 238U, 239Pu radial fission rate distributions the explanation of pickvalues discrepancies is based on the group fission constants analyses and possibleunderestimation of some features at the experimental data recovery method (Westcottfactors, temperature dependence).8:25 AMAdvanced Fuel Assembly Characterization Capabilities Basedon Gamma Tomography at the Halden Boiling Water ReactorScott Holcombe and Knut Eitrheim (1), Staffan Jacobsson Svärd (2), LarsHallstadius and Christofer Willman (3)1) Institute for Energy Technology, OECD Halden Reactor Project, Halden, Norway. 2) Division of AppliedNuclear Physics, Uppsala University, Uppsala, Sweden. 2) Westinghouse Electric Sweden AB, Västerås,SwedenCharacterization of individual fuel rods using gamma spectroscopy is a standard partof the Post Irradiation Examinations performed on experimental fuel at the HaldenBoiling Water Reactor. However, due to handling and radiological safety concerns,these measurements are presently carried out only at the end of life of the fuel, andnot earlier than several days or weeks after its removal from the reactor core. In orderto enhance the fuel characterization capabilities at the Halden facilities, a gamma tomographymeasurement system is now being constructed, capable of characterizingfuel assemblies on a rod-by-rod basis in a more timely and efficient manner. Gammatomography for measuring nuclear fuel is based on gamma spectroscopy measurementsand tomographic reconstruction techniques. The technique, previously demonstratedon irradiated commercial fuel assemblies, is capable of determining rod-byrodinformation without the need to dismantle the fuel. The new gamma tomographysystem will be stationed close to the Halden reactor in order to limit the need for fueltransport, and it will significantly reduce the time required to perform fuel characterizationmeasurements. Furthermore, it will allow rod-by-rod fuel characterization tooccur between irradiation cycles, thus allowing for measurement of experimental fuelrepeatedly during its irradiation lifetime. The development of the gamma tomographymeasurement system is a joint project between the Institute for Energy Technology –OECD Halden Reactor Project, Westinghouse (Sweden), and Uppsala University.9:15 AMExperimental Power Density Distribution Benchmark in TheTRIGA Mark II ReactorLuka Snoj, Žiga Štancar, Vladimir Radulovič, Manca Podvratnik, GašperŽerovnik, Andrej Trkov (1), Loic Barbot, Christophe Domergue, ChristopheDestouches (2)1) Jožef Stefan Institute, Ljubljana, Slovenia. 2) CEA, DEN, DER, Instrumentation Sensors and DosimetryLaboratory, Saint-Paul-Lez-Durance, FranceIn order to improve the power calibration process and to benchmark the existing computationalmodel of the TRIGA Mark II reactor at the Jožef Stefan Institute (JSI), a bilateralproject was started as part of the agreement between the French Commissariatà l’énergie atomique et aux énergies alternatives (CEA) and the Ministry of highereducation, science and technology of Slovenia. One of the objectives of the projectwas to analyze and improve the power calibration process of the JSI TRIGA reactor(procedural improvement and uncertainty reduction) by using absolutely calibratedCEA fission chambers (FCs). This is one of the few available power density distributionbenchmarks for testing not only the fission rate distribution but also the absolute valuesof the fission rates. Our preliminary calculations indicate that the total experimentaluncertainty of the measured reaction rate is sufficiently low that the experimentscould be considered as benchmark experiments.9:40 AMDevelopmoent of Triga-Based Experimental Device for FiberOptics In-Core Instrumentation Testing for VHTRsJesse M. Johns and Pavel V. TsvetkovDepartment of Nuclear Engineering, Texas A&M University, College Station, TXGiven the harsh environments of high temperature reactors, new in-core instrumentationhas to be developed, since existing approaches may fail prematurely in VHTRs.The paper discusses ongoing efforts to support progress of suitable advanced in-coreinstrumentation technologies and develop an experimental approach for evaluation oftheir performance within VHTRs via emulation of VHTR in-core conditions in TRIGAreactors. Successful completion of the presented computational analysis concludesthe first phase of the project. As demonstrated, it is proposed to use a high temperaturefurnace with fluence equivalency in operating TRIGA reactors.75

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 8:00 AM - 301 C14B - Reactor Transient & Safety AnalysisSession Chair: Barry Ganapol (UA); D. Diamond (BNL)8:00 AMQualification of CASMO5 / SIMULATE-3K Against the SPERT-III E-CORE Cold Start-Up ExperimentsGerardo Grandi and Lars MobergStudsvik Scandpower, Inc., Idaho Falls, ID, USASIMULATE-3K is a three-dimensional kinetic code applicable to LWR Reactivity InitiatedAccidents. S3K has been used to calculate several international recognizedbenchmarks. However, the feedback models in the benchmark exercises are differentfrom the feedback models that SIMULATE-3K uses for LWR reactors. For this reason,it is worth comparing the SIMULATE-3K capabilities for Reactivity Initiated Accidentsagainst kinetic experiments. The Special Power Excursion Reactor Test III was apressurized-water, nuclear-research facility constructed to analyze the reactor kineticbehavior under initial conditions similar to those of commercial LWRs. The SPERTIII E-core resembles a PWR in terms of fuel type, moderator, coolant flow rate, andsystem pressure. The initial test conditions (power, core flow, system pressure, coreinlet temperature) are representative of cold start-up, hot start-up, hot standby, and hotfull power. The qualification of S3K against the SPERT III E-core measurements is anongoing work at Studsvik. In this paper, the results for the 30 cold start-up tests arepresented. The results show good agreement with the experiments for the reactivityinitiated accident main parameters: peak power, energy release and compensatedreactivity. Predicted and measured peak powers differ at most by 13%. Measured andpredicted reactivity compensations at the time of the peak power differ less than 0.01$. Predicted and measured energy release differ at most by13%. All differences arewithin the experimental uncertainty.8:25 AMCoupled 3D-Neutronics / Thermal-Hydraulics Analysis of anUnprotected Loss-of-Flow Accident for a 3600 MWth SFRCoreKaichao Sun (1,2), Aurelia Chenu (2), Konstantin Mikityuk, Jiri Krepel (1),Rakesh Chawla (1,2)1) Paul Scherrer Institut (PSI), Villigen PSI, Switzerl. 2) Ecole Polytechnique Fédérale de Lausanne(EPFL), 1015 Lausanne, SwitzerlandThe core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigatedin this paper with the use of a coupled TRACE/PARCS model. The SFRneutron spectrum is characterized by several performance advantages, but also leadsto one dominating neutronics drawback – a positive sodium void reactivity. This impliesa positive reactivity effect when sodium coolant is removed from the core. In orderto evaluate such feedback in terms of the dynamics, a representative unprotectedloss-offlow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boilingoccurs, is analyzed. Although analysis of a single transient cannot allow generalconclusions to be drawn, it does allow better understanding of the underlying physicsand can lead to proposals for improving the core response during such an accident.The starting point of this study is the reference core design considered in the frameworkof the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR).To reduce the void effect, the core has been modified by introducing an upper sodiumplenum (along with a boron layer) and by reducing the core height-to-diameter ratio.For the ULOF considered, a sharp increase in core power results in melting of the fuelin the case of the reference core. In the modified core, a large dryout leads to meltingof the clad.8:50 AMEstimation of Average Burnup of Damaged Fuels Loaded inFukushima Dai-Ichi Reactors by Using the 134Cs/137Cs RatioMethodTomohiro Endo, Shunsuke Sato, and Akio YamamotoDepartment of Materials, Physics and Energy Engineering Graduate School of Engineering, NagoyaUniversity, Nagoya-shi, JapanAverage burnup of damaged fuels loaded in Fukushima Dai-ichi reactors is estimated,using the 134Cs/137Cs ratio method for measured radioactivities of 134Cs and 137Csin contaminated soils within the range of 100 km from the Fukushima Dai-ichi nuclearpower plants. As a result, the measured 134Cs/137Cs ratio from the contaminated soilis 0.996±0.07 as of March 11th, 2011. Based on the 134Cs/137Cs ratio method, theestimated burnup of damaged fuels is approximately 17.2±1.5 [GWd/tHM]. It is notedthat the numerical results of various calculation codes (SRAC2006/PIJ, SCALE6.0/TRITON, and MVP-BURN) are almost the same evaluation values of 134Cs/137Csratio with same evaluated nuclear data library (ENDF-B/VII.0). The void fraction effectin depletion calculation has a major impact on 134Cs/137Cs ratio compared with thedifferences between JENDL-4.0 and ENDF-B/VII.0.9:15 AMLevel 1 Transient Model for a Molybdenum-99 ProducingAqueous Homogeneous Reactor and its Applicability to theTracy ReactorE.T. Nygaard (1), M.M.R. Williams (2), P.L. Angelo (3)1) Babcock & Wilcox Technical Services Group, Lynchburg, VA. 2) Imperial College London, SW7 2AZ. 3)Y-12 National Security Complex, Oak Ridge, TNBabcock and Wilcox Technical Services Group (B&W) has identified aqueous homogeneousreactors (AHRs) as a technology well suited to produce the medical isotopemolybdenum 99 (Mo99). AHRs have never been specifically designed or built for thisspecialized purpose. However, AHRs have a proven history of being safe researchreactors. In fact, in 1958, AHRs had “a longer history of operation than any othertype of research reactor using enriched fuel” and had “experimentally demonstratedto be among the safest of all various type of research reactor now in use [1].” A “Level1” model representing B&W’s proposed Medical Isotope Production System (MIPS)reactor has been developed. The Level 1 model couples a series of differential equationsrepresenting neutronics, temperature, and voiding. Neutronics are representedby point reactor kinetics while temperature and voiding terms are axially varying (onedimensional).While this model was developed specifically for the MIPS reactor, itsapplicability to the Japanese TRACY reactor was assessed. The results from the Level1 model were in good agreement with TRACY experimental data and found to beconservative over most of the time domains considered. The Level 1 model was usedto study the MIPS reactor. An analysis showed the Level 1 model agreed well with amore complex computational model of the MIPS reactor (a FETCH model). Finally, asignificant reactivity insertion was simulated with the Level 1 model to study the MIPSreactor’s time-dependent response.9:40 AMSteps Towards Verification and Validation of the Fetch Codefor Level 2 Analysis, Design, and Optimization of AqueousHomogeneous ReactorsE.T. Nygaard (1), C.C. Pain, M.D. Eaton, J.L.M.A. Gomes, A.J.H. Goddard,G. Gorman, B. Tollit, A.G. Buchan, and C.M. Cooling (2), P.L. Angelo (3)1) Babcock & Wilcox Technical Services Group, Lynchburg VA. 2) Applied Modelling and ComputationGroup, Imperial College London. 3) Y-12 National Security Complex, Oak Ridge, TNBabcock and Wilcox Technical Services Group (B&W) has identified aqueous homogeneousreactors (AHRs) as a technology well suited to produce the medical isotopemolybdenum 99 (Mo99). AHRs have never been specifically designed or built for thisspecialized purpose. However, AHRs have a proven history of being safe researchreactors. In fact, in 1958, AHRs had “a longer history of operation than any other typeof research reactor using enriched fuel” and had “experimentally demonstrated to beamong the safest of all various type of research reactor now in use [1].” While AHRshave been modeled effectively using simplified “Level 1” tools, the complex interactionsbetween fluids, neutronics, and solid structures are important (but not necessarilysafety significant). These interactions require a “Level 2” modeling tool. ImperialCollege London (ICL) has developed such a tool: Finite Element Transient Criticality(FETCH). FETCH couples the radiation transport code EVENT with the computationalfluid dynamics code (Fluidity), the result is a code capable of modeling sub-critical,critical, and super-critical solutions in both twoand three-dimensions. Using FETCH,ICL researchers and B&W engineers have studied many fissioning solution systemsinclude the Tokimura criticality accident, the Y12 accident, SILENE, TRACY, andSUPO. These modeling efforts will ultimately be incorporated into FETCH’s extensiveautomated verification and validation (V&V) test suite expanding FETCH’s area of applicabilityto include all relevant physics associated with AHRs. These efforts parallelB&W’s engineering effort to design and optimize an AHR to produce Mo99.76

8:00 AMAxial Grading of Inert Matrix FuelsGeoffrey D. Recktenwald and Mark R. DeinertDepartment of Mechanical Engineering University of Texas at AustinPHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 8:00 AM - 301 D7D - Fuel Cycle & Actinide ManagementSession Chair: Nick Touran (TerraPower); Sara Bortot (KTH)Burning actinides in an inert matrix fuel to 750 MWd/kgIHM results in a significant reductionin transuranic isotopes. However, achieving this level of burnup in a standardlight water reactor would require residence times that are twice that of uranium dioxidefuels. The reactivity of an inert matrix assembly at the end of life is less than 1/3 of itsbeginning of life reactivity leading to undesirable radial and axial power peaking in thereactor core. Here we show that axial grading of the inert matrix fuel rods can reducepeaking significantly. Montecarlo simulations are used to model the assembly levelpower distributions in both ungraded and graded fuel rods. The results show that anaxial grading of uranium dioxide and inert matrix fuels with erbium can reduces powerpeaking by more than 50% in the axial direction. The reduction in power peaking enablesthe core to operate at significantly higher power.8:25 AMPlutonium and Minor Actinide Utilisation in a Pebble-Bed HighTemperature ReactorB.Y. Petrov, J.C. Kuijper, J. Oppe, J.B.M. de HaasNuclear Research and Consultancy Group Westerduinweg 3, Petten, The NetherlandsThis paper contains results of the analysis of the pebble-bed high temperature gascooledPUMA reactor loaded with plutonium and minor actinide (Pu/MA) fuel. Startingfrom knowledge and experience gained in the Euratom FP5 projects HTR-N andHTR-N1, this study aims at demonstrating the potential of high temperature reactorsto utilize or transmute Pu/MA fuel. The work has been performed within the EuratomFP6 project PUMA. A number of different fuel types and fuel configurations have beenanalyzed and compared with respect to incineration performance and safety-relatedreactor parameters. The results show the excellent plutonium and minor actinide burningcapabilities of the high temperature reactor. The largest degree of incineration isattained in the case of an HTR fuelled by pure plutonium fuel as it remains critical atvery deep burnup of the discharged pebbles. Addition of minor actinides to the fuelleads to decrease of the achievable discharge burnup and therefore smaller fraction ofactinides incinerated during reactor operation. The inert-matrix fuel design improvesthe transmutation performance of the reactor, while the “wallpaper” fuel does not haveadvantage over the standard fuel design in this respect. After 100 years of decay followingthe fuel discharge, the total amount of actinides remains almost unchangedfor all of the fuel types considered. Among the plutonium isotopes, only the amount ofPu-241 is reduced significantly due to its relatively short half-life.8:50 AMDual Neutral Particle Transmutation in CINDER2008W. J. Martin and C. R. E. de OliveiraUniversity of New Mexico, Albuquerque, NMA capability has been built for the CINDER2008 (beta) transmutation code that expandsthe capability from only neutron induced reactions to photon induced reactions.This allows for two incident neutral particles to cause nuclear transmutation in a givenmaterial simultaneously. The CINDER2008 code, a modular rewrite of the CINDER’90transmutation code from Los Alamos National Laboratory, was modified to allow forthe dual sets of physics. A photonuclear cross section and photofission product yieldlibrary was also created using ENDF-B/VII data and translated neutron fission productyields. The code and library have been combined to create a unique transmutationcode. The scope of use is broad; it is capable of modeling the transmutation causedby photons released from the decay of daughter and fission products as well as transmutationin photon rich environments. A brief code description and a verification andvalidation of the contributions are given.9:15 AMModeling Depletion Simulations for a High-Burnup, HighlyHeterogeneous BWR Fuel Assembly with ScaleH. J. SmithOak Ridge National Laboratory, Oak Ridge, TNExtensive SCALE isotopic validation studies have been performed for various PWRfuel assembly designs and operating conditions, and to a lesser extent for BWR fuelassembly designs. However, no SCALE validation work has been documented fornewer, highly heterogeneous BWR fuel assembly designs at high burnup. Isotopicbenchmark calculations of the earlier, more geometrically uniform BWR fuel assembliesare less sensitive to simplification of the operating history details and certainmodeling assumptions than heterogeneous fuel assemblies, particularly at high burnup.This analysis shows the capability of SCALE to simulate a complex highly heterogeneousSVEA96 Optima fuel assembly and illustrates the importance of the needfor the highest possible accuracy and precision in isotope measurements intended tobe used as benchmark-quality results. In addition, this analysis quantifies the impactof various modeling assumptions on the results. The sample for which the simulationresults are reported here achieved a burnup 62 GWd/MTU and was analyzed as partof the MALIBU Extension program.9:40 AMShort-Time Scale Behavior Modeling within Long-Time ScaleFuel Cycle EvaluationsM. Johnson, P. Tsvetkov (1), S. Lucas (2)1) Department of Nuclear Engineering, Texas A&M University, College Station, TX. 2) Idaho NationalLaboratory, Idaho Falls, IdahoTypically, short-time and long-time scales in nuclear energy system behavior are accountedfor with entirely separate models. However, long-term changes in systemcharacteristics do affect short-term transients through material variations. This paperpresents an approach to consistently account for short-time scales within a nuclearsystem lifespan. The reported findings and developments are of significant importancefor small modular reactors and other nuclear energy systems operating in autonomousmodes. It is necessary to simulate the short time-scale kinetic behavior of the reactoras well as the long time-scale dynamics that occur with fuel burnup. The former ismodeled using the point kinetics equations, while the latter is modeled by the Batemanequations.77

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 8:00 AM - 301 E8E - Advanced Modeling & Simulation in Reactor PhysicsSession Chair: Tom Sutton (KAPL); K. Ivanov (PSU)8:00 AMModeling CANDU-6 Liquid Zone Controllers for Effects ofThorium-Based FuelsEmmanuel St-Aubin and Guy MarleauÉcole Polytechnique de Montréal, Montréal (QC), CanadaWe use the DRAGON code to model the CANDU-6 liquid zone controllers and evaluatethe effects of thorium-based fuels on their incremental cross sections and reactivityworth. We optimize both the numerical quadrature and spatial discretization for 2Dcell models in order to provide accurate fuel properties for 3D liquid zone controllersupercell models. We propose a low computer cost parameterized pseudo-exact 3Dcluster geometries modeling approach that avoids tracking issues on small externalsurfaces. This methodology provides consistent incremental cross sections and reactivityworths when the thickness of the buffer region is reduced. When compared withan approximate annular geometry representation of the fuel and coolant region, weobserve that the cluster description of fuel bundles in the supercell models does notincrease considerably the precision of the results while increasing substantially theCPU time. In addition, this comparison shows that it is imperative to finely describe theliquid zone controller geometry since it has a strong impact of the incremental crosssections. This paper also shows that liquid zone controller reactivity worth is greatlydecreased in presence of thorium-based fuels compared to the reference natural uraniumfuel, since the fission and the fast to thermal scattering incremental cross sectionsare higher for the new fuels.8:25 AMSpecification of Requirements for the Virtual Environment forReactor Applications Simulation EnvironmentStephen M. Hess (1), Martin Pytel (2)1) Electric Power Research Institute, West Chester, PA, USA. 2) Electric Power Research Institute, PaloAlto, CA, USAIn 2010, the United States Department of Energy initiated a research and developmenteffort to develop modern modeling and simulation methods that could utilizehigh performance computing capabilities to address issues important to nuclear powerplant operation, safety and sustainability. To respond to this need, a consortium ofnational laboratories, academic institutions and industry partners (the Consortium forAdvanced Simulation of Light Water Reactors – CASL) was formed to develop anintegrated Virtual Environment for Reactor Applications (VERA) modeling and simulationcapability. A critical element for the success of the CASL research and developmenteffort was the development of an integrated set of overarching requirementsthat provides guidance in the planning, development, and management of the VERAmodeling and simulation software. These requirements also provide a mechanismfrom which the needs of a broad array of external CASL stakeholders (e.g. reactor /fuel vendors, plant owner / operators, regulatory personnel, etc.) can be identified andintegrated into the VERA development plans. This paper presents an overview of theinitial set of requirements contained within the VERA Requirements Document (VRD)that currently is being used to govern development of the VERA software within theCASL program. The complex interdisciplinary nature of these requirements togetherwith a multi-physics coupling approach to realize a core simulator capability pose achallenge to how the VRD should be derived and subsequently revised to accommodatethe needs of different stakeholders. Thus, the VRD is viewed as an evolvingdocument that will be updated periodically to reflect the changing needs of identifiedCASL stakeholders and lessons learned during the progress of the CASL modelingand simulation program.8:50 AMStability and Convergence Analysis of the Quasi- DynamicsMethod for the Initial Pebble PackingYanheng Li and Wei JiDepartment of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy,NYThe simulation for the pebble flow recirculation within Pebble Bed Reactors (PBRs)requires an efficient algorithm to generate an initial overlap-free pebble configurationwithin the reactor core. In the previous work, a dynamics-based approach, the Quasi-Dynamics Method (QDM), has been proposed to generate densely distributed pebblesin PBRs with cylindrical and annular core geometries. However, the stability and theefficiency of the QDM were not fully addressed. In this work, the algorithm is reformulatedwith two control parameters and the impact of these parameters on the algorithmperformance is investigated. Firstly, the theoretical analysis for a 1-D packing systemis conducted and the range of the parameter in which the algorithm is convergentis estimated. Then, this estimation is verified numerically for a 3-D packing system.Finally, the algorithm is applied to modeling the PBR fuel loading configuration and theconvergence performance at different packing fractions is presented. Results showthat the QDM is efficient in packing pebbles within the realistic range of the packingfraction in PBRs, and it is capable in handling cylindrical geometry with packing fractionsup to 63.5%.9:15 AMSimulation of a Main Steam Line Break with Steam GeneratorTube Rupture Using TraceS. Gallardo, A. Querol and G. VerdúDepartamento de Ingeniería Química y Nuclear, Universitat Politècnica de València, València, SpainA simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermalhydrauliccode TRACE5. Test 5 performed in the Large Scale Test Facility (LSTF)reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture(SGTR) in a Pressurized Water Reactor (PWR). The result of these simultaneousbreaks is a depressurization in the secondary and primary system in loop B becauseboth systems are connected through the SGTR. Good approximation was obtainedbetween TRACE5 results and experimental data. TRACE5 reproduces qualitativelythe phenomena that occur in this transient: primary pressure falls after the break,stagnation of the pressure after the opening of the relief valve of the intact steam generator,the pressure falls after the two openings of the PORV and the recovery of theliquid level in the pressurizer after each closure of the PORV. Furthermore, a sensitivityanalysis has been performed to know the effect of varying the High Pressure Injection(HPI) flow rate in both loops on the system pressures evolution.9:40 AMConvergence Analysis of a CMFD Method Based on GeneralizedEquivalence TheoryYunlin Xu (1), T. Downar (2)1) Argonne National Laboratory, Argonne, IL. 2) Department of Nuclear Engineering and RadiologicalSciences, University of Michigan, Ann Arbor, MICMFD acceleration methods have been successful in reducing the computationalburden for steady-state and transient reactor calculations. However, recent work ona complex coupled code BWR ATWS event has exposed possible issues with thestability of the CMFD method when standard CMFD methods are used. During thesimulation of the ATWS boron injection event in the BWR, the PARCS code failed toconverge with the existing CMFD method. A new CMFD method based on generalizedequivalence theory was developed and the PARCS solution converged for thesame ATWS event. This paper presents the new method and a detailed analytic andnumerical convergence analysis. The results show that the new CMFD converges forall possible cross sections combinations anticipated in Light Water Reactor simulationand unlike existing CMFD methods, it is very robust even when the initial guess is farfrom final true solution.78

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 10:20 AM - 301 A1G - Core Analysis MethodsSession Chair: Cristian Rabiti (INL)10:20 AMInvestigation of the MTC noise estimation with a coupledneutronic/thermal-hydraulic dedicated model – “Closing theloop”Christophe Demazière and Viktor LarssontChalmers University of Technology, Gothenburg, SwedenThis paper investigates the reliability of different noise estimators aimed at determiningthe Moderator Temperature Coefficient (MTC) of reactivity in Pressurized WaterReactors. By monitoring the inherent fluctuations in the neutron flux and moderatortemperature, an on-line monitoring of the MTC without perturbing reactor operationis possible. In order to get an accurate estimation of the MTC by noise analysis, thepoint-kinetic component of the neutron noise and the core-averaged moderator temperaturenoise have to be used. Because of the scarcity of the in-core instrumentation,the determination of these quantities is difficult, and several possibilities thus existfor estimating the MTC by noise analysis. Furthermore, the effect of feedback has tobe negligible at the frequency chosen for estimating the MTC in order to get a properdetermination of the MTC. By using an integrated neutronic/thermal-hydraulic modelspecifically developed for estimating the three-dimensional distributions of the fluctuationsin neutron flux, moderator properties, and fuel temperature, different approachesfor estimating the MTC by noise analysis can be tested individually. It is demonstratedthat a reliable MTC estimation can only be provided if the core is equipped with asufficient number of both neutron detectors and temperature sensors, i.e. if the corecontain in-core detectors monitoring both the axial and radial distributions of the fluctuationsin neutron flux and moderator temperature. It is further proven that the effectof feedback is negligible for frequencies higher than 0.1 Hz, and thus the MTC noiseestimations have to be performed at higher frequencies.10:45 AMUnderstanding the Haling Power Depletion (HPD) MethodS. Levine, T. Blyth, and K. IvanovThe Pennsylvania State University (PSU), University Park, PA, USAThe Pennsylvania State University (PSU) is using the university version of the StudsvikScandpower Code System (CMS) for research and education purposes. Preparationshave been made to incorporate the CMS into the PSU Nuclear Engineeringgraduate class “Nuclear Fuel Management” course. The information presented in thispaper was developed during the preparation of the material for the course. The HalingPower Depletion (HPD) was presented in the course for the first time. The HPDmethod has been criticized as not valid by many in the field even though it has beensuccessfully applied at PSU for the past 20 years. It was noticed that the radial powerdistribution (RPD) for low leakage cores during depletion remained similar to that ofthe HPD during most of the cycle. Thus, the Haling Power Depletion (HPD) may beused conveniently mainly for low leakage cores. Studies were then made to betterunderstand the HPD and the results are presented in this paper. Many different coreconfigurations can be computed quickly with the HPD without using Burnable Poisons(BP) to produce several excellent low leakage core configurations that are viable forpower production. Once the HPD core configuration is chosen for further analysis,techniques are available for establishing the BP design to prevent violating any of thesafety constraints in such HPD calculated cores. In summary, in this paper it has beenshown that the HPD method can be used for guiding the design for the low leakagecore.11:10 AMBenchmark Problems and Results for Verifying ResonanceCalculation MethodologiesHongchun Wu, Weiyan Yang, Yulong Qin, Lei He, Liangzhi Cao, YouqiZheng, and Qingjie LiuNECP laboratory, School of Nuclear Science and Technology, Xi’an Jiaotong University, ChinaResonance calculation is one of the most important procedures for the multi-groupneutron transport calculation. With the development of nuclear reactor concepts, manynew types of fuel assembly are raised. Compared to the traditional designs, most ofthe new fuel assemblies have different fuel types either with complex isotopes or withcomplicated geometry. This makes the traditional resonance calculation method invalid.Recently, many advanced resonance calculation methods are proposed. However,there are few benchmark problems for evaluating those methods with a comprehensivecomparison. In this paper, we design 5 groups of benchmark problems including21 typical cases of different geometries and fuel contents. The reference results of thebenchmark problems are generated based on the sub-group method, ultra-fine groupmethod, function expanding method and Monte Carlo method. It is shown that thosebenchmark problems and their results could be helpful to evaluate the validity of thenewly developed resonance calculation method in the future work.11:35 AMTowards a Reference Numerical Scheme Using MCNPX forPWR Control Rod Tip Fluence EstimationsHakim Ferroukhi, Alexander Vasiliev (1), Alice Dufresne, Rakesh Chawla(2)1) Paul Scherrer Institut, Villigen-PSI, Switzerland. 2) Department of Physics, EPFL, Lausanne, SwitzerlandRecent occurrences of cracks and fissures on the cladding tubes of PWR control rod(CR) fingers employed in the Swiss reactors prompted the need to develop more reliableanalytical methods for CR tip fluence estimations. To partly address this need, adeterministic methodology based on SIMULATE-3/CASMO-4 was in recent years developedat PSI. Although this methodology has already been applied for independentsupport to licensing issues related to CR lifetime, two main questions are currentlybeing the center of attention for further enhancements. First, the methodology relieson several assumptions that have so far not been verified. Secondly, an assessmentof the achieved accuracy has not been addressed. In an attempt to answer both theseopen questions, it was considered appropriate to develop an alternative computationalscheme based on the stochastic MCNPX code with the objective to provide referencenumerical solutions. This paper presents the first steps undertaken in that direction.To start, a methodology for a volumetric neutron source transfer to full core MCNPXmodels with detailed CR as well as axial reflector representations is established. Onthis basis, the assumptions of the deterministic methodology are studied for selectedCR configurations for two Beginning-of-Life cores by comparing the spatial neutronflux distributions obtained with the two approaches for the entire spectrum. Finally, forthe high-energy range (E> 1 MeV) and for a few CRs, the new MCNPX scheme is appliedto estimate the accumulated fluence over one real operated cycle and the resultsare compared with the deterministic approach.79

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 10:20 AM - 301 B12C - Sensitivity & Uncertainty AnalysisSession Chair: A. Santamarina (CEA); W. Zwermann (GRS)10:20 AMA Surrogate-Based Uncertainty Quantification with QuantifiableErrorsYoungsuk Bang and Hany S. Abdel-KhalikNorth Carolina State University, Raleigh, NCSurrogate models are often employed to reduce the computational cost requiredto complete uncertainty quantification, where one is interested in propagating inputparameters uncertainties throughout a complex engineering model to estimate responsesuncertainties. An improved surrogate construction approach is introducedhere which places a premium on reducing the associated computational cost. Unlikeexisting methods where the surrogate is constructed first, then employed to propagateuncertainties, the new approach combines both sensitivity and uncertainty informationto render further reduction in the computational cost. Mathematically, the reduction isdescribed by a range finding algorithm that identifies a subspace in the parametersspace, whereby parameters uncertainties orthogonal to the subspace contribute negligibleamount to the propagated uncertainties. Moreover, the error resulting from thereduction can be upper-bounded. The new approach is demonstrated using a realisticnuclear assembly model and compared to existing methods in terms of computationalcost and accuracy of uncertainties. Although we believe the algorithm is general, it willbe applied here for linear-based surrogates and Gaussian parameters uncertainties.The generalization to nonlinear models will be detailed in a separate article.10:45 AMLow Rank Approach to Computing First and Higher Order DerivativesUsing Automatic DifferentiatonJames A. Reed and Hany S. Abdel-Khalik (1), Jean Utke (2)1) North Carolina State University, Department of Nuclear Engineering, Raleigh, NC. 2) Mathematics andComputer Science Division, Argonne National Laboratory, Argonne, ILThis manuscript outlines a new approach for increasing the efficiency of applying automaticdifferentiation (AD) to large scale computational models. By using the principlesof the Efficient Subspace Method (ESM), low rank approximations of the derivativesfor first and higher orders can be calculated using minimized computational resources.The output obtained from nuclear reactor calculations typically has a much smallernumerical rank compared to the number of inputs and outputs. This rank deficiencycan be exploited to reduce the number of derivatives that need to be calculated usingAD. The effective rank can be determined according to ESM by computing derivativeswith AD at random inputs. Reduced or pseudo variables are then defined andnew derivatives are calculated with respect to the pseudo variables. Two differentAD packages are used: OpenAD and Rapsodia. OpenAD is used to determine theeffective rank and the subspace that contains the derivatives. Rapsodia is then usedto calculate derivatives with respect to the pseudo variables for the desired order. Theoverall approach is applied to two simple problems and to MATWS, a safety code forsodium cooled reactors.11:10 AMAN EpGPT-Based Approach for Uncertainty QuantificationCongjian Wang and Hany S. Abdel-KhalikDepartment of Nuclear Engineering, North Caroline State University, Raleigh, NCGeneralized Perturbation Theory (GPT) has been widely used by many scientific disciplinesto perform sensitivity analysis and uncertainty quantification. This manuscriptemploys recent developments in GPT theory, collectively referred to as Exact-to-PrecisionGeneralized Perturbation Theory (EPGPT), to enable uncertainty quantificationfor computationally challenging models, e.g. nonlinear models associated with manyinput parameters and many output responses and with general non-Gaussian parametersdistributions. The core difference between EPGPT and existing GPT is in the waythe problem is formulated. GPT formulates an adjoint problem that is dependent on theresponse of interest. It tries to capture via the adjoint solution the relationship betweenthe response of interest and the constraints on the state variations. EPGPT recaststhe problem in terms of a smaller set of what is referred to as the ‘active’ responseswhich are solely dependent on the physics model and the boundary and initial conditionsrather than on the responses of interest. The objective of this work is to apply anEPGPT methodology to propagate cross-sections variations in typical reactor designcalculations. The goal is to illustrate its use and the associated impact for situationswhere the typical Gaussian assumption for parameters uncertainties is not valid andwhen nonlinear behavior must be considered. To allow this demonstration, exaggeratedvariations will be employed to stimulate nonlinear behavior in simple prototypicalneutronics models.11:35 AMStatistical Uncertainty Analisis Applied to the Dragonv4 CodeLattice Calculations and Based on JENDL-4 Covariance DataAugusto Hernandez-Solis (1), Christophe Demazière (2), Christian Ekberg,Arvid Ödegaard-Jensen (1)1) Department of Nuclear Chemistry, Chalmers University of Technology, Göteborg, Sweden. 2) Departmentof Nuclear Engineering, Chalmers University of Technology, Göteborg, SwedenIn this paper, multi-group microscopic cross-section uncertainty is propagated throughthe DRAGON (Version 4) lattice code, in order to perform uncertainty analysis on ...and 2-group homogenized macroscopic cross-sections predictions. A statistical methodologyis employed for such purposes, where cross-sections of certain isotopes ofvarious elements belonging to the 172 groups DRAGLIB library format, are consideredas normal random variables. This library is based on JENDL-4 data, because JENDL-4contains the largest amount of isotopic covariance matrixes among the different majornuclear data libraries. The aim is to propagate multi-group nuclide uncertainty byrunning the DRAGONv4 code 500 times, and to assess the output uncertainty of atest case corresponding to a 17x17 PWR fuel assembly segment without poison. Thechosen sampling strategy for the current study is Latin Hypercube Sampling (LHS).The quasi-random LHS allows a much better coverage of the input uncertainties thansimple random sampling (SRS) because it densely stratifies across the range of eachinput probability distribution. Output uncertainty assessment is based on the tolerancelimits concept, where the sample formed by the code calculations infers to cover 95%of the output population with at least a 95% of confidence. This analysis is the first attemptto propagate parameter uncertainties of modern multi-group libraries, which areused to feed advanced lattice codes that perform state of the art resonant selfshieldingcalculations such as DRAGONv4.80

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 10:20 AM - 301 C3F - Monte Carlo Methods & DevelopmentsSession Chair: Forrest Brown (LANL); Tom Sutton (KAPL)10:20 AMImpact of Delayed Neutron Precursor Mobility in Fissile SolutionSystemsBrian C. KiedrowskiX-Computational Physics Division, Los Alamos National Laboratory, Los Alamos, NM11:10 AMNeutron Source in the MCNPX Shielding Calculating for ElectronAccelerator Driven FacilityZhaopeng Zhong, Yousry GoharNuclear Engineering Division, Argonne National Laboratory, Argonne, ILA research version of the Monte Carlo software package MCNP6 is modified to incorporateadvection and diffusion of delayed neutron precursors, resulting in the emissionof delayed neutrons at locations different from the original fission sites. Results of twotest problems, a pipe carrying flowing fissile solution and a sphere of fissile solutionwith precursor diffusion, show that the fission product mobility tends to perturb thefundamental mode, has a negative reactivity effect, and, perhaps most importantly,causes a decrease in the effective delayed neutron fraction.10:45 AMChord Length Sampling Method for Analyzing VHTR Unit Cellsin Continuous Energy SimulationsChao Liang and Wei Ji (1), Forrest B. Brown (2)1) Department of Mechanical, Aerospace, and Nuclear Engineering Rensselaer Polytechnic Institute,Troy, NY. 2) Los Alamos National Laboratory, Los Alamos NMThe chord length sampling method (CLS) is studied in the continuous energy simulationsby applying it to analyzing two types of Very High Temperature Gas-cooled Reactor(VHTR) unit cells: the fuel compact cell in the prismatic type VHTR and the fuelpebble cell in the pebble-bed type VHTR. Infinite multiplication factors of the unit cellsare calculated by the CLS and compared to the benchmark simulations at differentvolume packing fractions from 5% to 30%. It is shown that the accuracy of the CLS isaffected by the boundary effect, which is induced by the CLS procedure itself and resultsin a reduction in the total volume packing fraction of the fuel particles. To mitigatethe boundary effect, three correction schemes based on the research of 1) Murata etal. 2) Ji and Martin 3) Griesheimer et al. are used to improve the accuracy by applyinga corrected value of the volume packing fraction to the CLS. These corrected valuesare calculated based on 1) a simple linear relationship, 2) an iterative self-consistentsimulation correction method, and 3) a theoretically derived non-linear relationship,respectively. The CLS simulation using the corrected volume packing fraction showsexcellent improvements in the infinite multiplication factors for the VHTR unit cells. Jiand Martin’s self-consistent correction method shows the best improvement.Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology(KIPT) of Ukraine have been collaborating on the design development of anexperimental neutron source facility. It is an accelerator driven system (ADS) utilizinga subcritical assembly driven by electron accelerator. The facility will be utilized for performingbasic and applied nuclear researches, producing medical isotopes, and trainingyoung nuclear specialists. Monte Carlo code MCNPX has been utilized as a designtool due to its capability to transport electrons, photons, and neutrons at high energies.However the facility shielding calculations with MCNPX need enormous computationalresources and the small neutron yield per electron makes sampling difficulty for theMonte Carlo calculations. A method, based on generating and utilizing neutron sourcefile, was proposed and tested. This method reduces significantly the required computerresources and improves the statistics of the calculated neutron dose outside theshield boundary. However the statistical errors introduced by generating the neutronsource were not directly represented in the results, questioning the validity of thismethodology, because an insufficiently sampled neutron source can cause error onthe calculated neutron dose. This paper presents a procedure for the validation of thegenerated neutron source file. The impact of neutron source statistic on the neutrondose is examined by calculating the neutron dose as a function of the number ofelectron particles used for generating the neutron source files. When the value of thecalculated neutron dose converges, it means the neutron source has scored sufficientrecords and statistic does not have apparent impact on the calculated neutron dose. Inthis way, the validity of neutron source and the shield analyses could be verified.11:35 AMA Monte Carlo Neutron Transport Code for Eigenvalue Calculationson a Dual-Gpu System and CUDA EnvironmentTianyu Liu, Aiping Ding, Wei Ji, and X. George Xu (1), Christopher D.Carothers (2), Forrest B. Brown (3)1) Nuclear Engineering and Engineering Physics, Rensselaer Polytechnic Institute, Troy, NY. 2) Departmentof Computer Science, Rensselaer Polytechnic Institute (RPI). 3) Los Alamos National Laboratory(LANL)Monte Carlo (MC) method is able to accurately calculate eigenvalues in reactor analysis.Its lengthy computation time can be reduced by general-purpose computing onGraphics Processing Units (GPU), one of the latest parallel computing techniquesunder development. The method of porting a regular transport code to GPU is usuallyvery straightforward due to the “embarrassingly parallel” nature of MC code. However,the situation becomes different for eigenvalue calculation in that it will be performedon a generation-by-generation basis and the thread coordination should be explicitlytaken care of. This paper presents our effort to develop such a GPU-based MC codein Compute Unified Device Architecture (CUDA) environment. The code is able toperform eigenvalue calculation under simple geometries on a multi-GPU system. Thespecifics of algorithm design, including thread organization and memory managementwere described in detail. The original CPU version of the code was tested on an IntelXeon X5660 2.8GHz CPU, and the adapted GPU version was tested on NVIDIA TeslaM2090 GPUs. Double-precision floating point format was used throughout the calculation.The result showed that a speedup of 7.0 and 33.3 were obtained for a barespherical core and a binary slab system respectively. The speedup factor was furtherincreased by a factor of ~2 on a dual GPU system. The upper limit of device-levelparallelism was analyzed, and a possible method to enhance the thread-level parallelismwas proposed.81

Session Chair: Mike Dunn (ORNL)PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 10:20 AM - 301 D10:20 AMThermal Neutron Capture Cross Section of Gadolinium byPile-Oscillation Measurments in MinervePierre Leconte, Jacques Di-Salvo, Muriel Antony, Alexandra Pepino (1), AymenHentati (2)1) CEA, DEN, DER, Cadarache, Saint-Paul-Lez-Durance. 2) International School in Nuclear Engineering,Cadarache, Saint-Paul-Lez-DuranceNatural gadolinium is used as a burnable poison in most LWR to account for the excessof reactivity of fresh fuels. For an accurate prediction of the cycle length, itsnuclear data and especially its neutron capture cross section needs to be known with ahigh precision. Recent microscopic measurements at Rensselaer Polytechnic Institute(RPI) suggest a 11% smaller value for the thermal capture cross section of 157Gd,compared with most of evaluated nuclear data libraries. To solve this inconsistency, wehave analyzed several pile-oscillation experiments, performed in the MINERVE reactor.They consist in the measurement of the reactivity variation involved by the introductionin the reactor of small-samples, containing different mass amounts of naturalgadolinium. The analysis of these experiments is done through the exact perturbationtheory, using the PIMS calculation tool, in order to link the reactivity effect to the thermalcapture cross section. The measurement of reactivity effects is used to deducethe 2200 m.s-1 capture cross section of natGd which is (49360 ± 790) b. This resultis in good agreement with the JEFF3.1.1 value (48630 b), within 1.6% uncertainty at1s, but is strongly inconsistent with the microscopic measurements at RPI which give(44200 ± 500) b.10:45 AMMICROX-2 Cross Section Library Based On ENDF/B-VIIJia Hou and Kostadin Ivanov (1), Hangbok Choi (2)Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park,PA. 2) General Atomics, San Diego, CANew cross section libraries of a neutron transport code MICROX-2 have been generatedfor advanced reactor design and fuel cycle analyses. A total of 386 nuclides wereprocessed, including 10 thermal scattering nuclides, which are available in ENDF/B-VIIrelease 0 nuclear data. The NJOY system and MICROR code were used to processnuclear data and convert them into MICROX-2 format. The energy group structure ofthe new library was optimized for both the thermal and fast neutron spectrum reactorsbased on Contributon and Point-wise Cross Section Driven (CPXSD) method, resultingin a total of 1173 energy groups. A series of lattice cell level benchmark calculationshave been performed against both experimental measurements and Monte Carlocalculations for the effective/infinite multiplication factor and reaction rate ratios. Theresults of MICROX-2 calculation with the new library were consistent with those of 15reference cases. The average errors of the infinite multiplication factor and reactionrate ratio were 0.31% k and 1.9%, respectively. The maximum error of reaction rateratio was 8% for 238U-to-235U fission of ZEBRA lattice against the reference calculationdone by MCNP5.11C - Nuclear Data11:10 AMCASMO5 JENDL-4.0 and ENDF/B-VII.1beta4 LibrariesJoel Rhodes, Nicholas Gheorghiu, and Rodolfo FerrerStudsvik Scandpower, Inc., Idaho Falls, IDThis paper details the generation of neutron data libraries for the CASMO5 latticephysics code based on the recently released JENDL-4.0 and ENDF/B-VII.1beta4 nucleardata evaluations. This data represents state-of-the-art nuclear data for late-2011.The key features of the new evaluations are briefly described along with the procedurefor processing of this data into CASMO5, 586- energy group neutron data libraries.Finally some CASMO5 results for standard UO2 and MOX critical experiments forthe two new libraries and the current ENDF/B-VII.0 CASMO5 library are presentedincluding the B&W 1810 series, DIMPLE S06A, S06B, TCA reflector criticals with ironplates and the PNL-30-35 MOX criticals. The results show that CASMO5 with the newlibraries is performing well for these criticals with a very slight edge in results to theJENDL-4.0 nuclear data evaluation over the ENDF/B-VII.1beta4 evaluation. Work iscurrently underway to generate a CASMO5 library based on the final ENDF/B-VII.R1evaluation released Dec. 22, 2011.11:35 AMValidating The ENDF-B/VII 235U(nth,f) Prompt Fission NeutronSpectrum Using Updated Dosimetry Cross Sections(IRDFF)Roberto Capote (1), Konstantin I. Zolotarev and Vladimir G. Pronyaev (2),Andrej Trkov (3)1) International Atomic Energy Agency, NAPC-Nuclear Data Section, Vienna, Austria. 2) Institute ofPhysics and Power Engineering, Obninsk, Kaluga Region, Russia. 3) Jozef Stefan Institute, Ljubljana,SloveniaThe International Reactor Dosimetry File IRDF-2002 released in 2004 by the IAEAcontains cross-section data and corresponding uncertainties for 66 dosimetry reactions.New cross-section evaluations have become available recently that re-definesome of these dosimetry reactions for reactor applications including: 1) high fidelityevaluation work undertaken by one of the authors (KIZ); 2) evaluations from theENDF/B-VII libraries that cover reactions within the International Evaluation of NeutronCross-Section Standards; and 3) evaluations from JENDL-3.1 and JENDL- 4 libraries.Overall, 37 new evaluations of dosimetry reactions have been assessed to determinewhether they should be adopted to update and improve IRDF-2002. A new dosimetrylibrary (International Reactor Dosimetry File for Fission and Fusion - IRDFF) was assembledbased on new evaluations combined with selected IRDF-2002 evaluations.A grand-total of 74 dosimetry reactions are included into the IRDFF dosimetry libraryavailable at www-nds.iaea.org/IRDFF1. The assembled library was used to validatethe 235U(nth,f) ENDF-B/VII.0 prompt fission neutron spectrum. An excellent averageC/E value of 1.002 +/- 0.02 is achieved for reactions with mean neutron energy of theintegrated response (E50%) lower than 11 MeV. C/E data for reactions with E50%-response higher than 11 MeV decreases up to 0.8. We conclude that the ENDF-B/VII.0235U(nth,f) prompt fission neutron spectrum from 1-11 MeV is validated within quoteduncertainties by available integral measurements in 235U(nth,f) neutron field. Furtherinvestigations for high-threshold reactions are needed and new measurements ofspectrum average cross sections for those reactions in the 235U(nth,f) neutron fieldare recommended.82

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 10:20 AM - 301 E8F - Advanced Modeling & Simulation in Reactor PhysicsSession Chair: Sandra Dulla (PdiT); Sara Bortot (KPH)10:20 AMIntegrated Radiation Transport and Nuclear Fuel Performancefor Assembly-Level SimulationsSteven Hamilton, Kevin Clarno, Bobby Philip, Mark Berrill, Rahul Sampathand Srikanth AlluOak Ridge National Laboratory, Oak Ridge, TNThe Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) isfocused on predicting the temperature and strain within a nuclear fuel assembly toevaluate the performance of existing and advanced nuclear fuel bundles within nuclearreactors. AMPFuel was extended to include an integrated nuclear fuel assembly capabilityfor (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics.This capability is the initial step toward incorporating an improved predictivenuclear fuel assembly modeling capability to accurately account for source-terms (i.e.neutron flux distribution, coolant conditions, and assembly mechanical stresses) of traditionalnuclear fuel performance simulations. A novel scheme is introduced for transferringthe power distribution from the SCALE/Denovo (Denovo) radiation transportcode (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel(unstructured, hexagonal mesh with a single material within each cell), allowing theuse of a relatively coarse spatial mesh for the radiation transport and a fine spatialmesh for thermo-mechanics with very little loss of accuracy. With this new capability,AMPFuel was used to model an entire 1717 pressurized water reactor fuel assemblywith many of the features resolved in three dimensions (for thermo-mechanics and/orneutronics). A full assembly calculation was executed on Jaguar using 40,000 coresin under 10 hours to model over 160 billion degrees of freedom for 10 loading steps.The single radiation transport calculation required about 50% of the time required tosolve the thermo-mechanics with a single loading step, which demonstrates that it isfeasible to incorporate, in a single code, a high-fidelity radiation transport capabilitywith a high-fidelity nuclear fuel thermo-mechanics capability and anticipate acceptablecomputational requirements. The results of the full assembly simulation clearly showthe axial, radial, and azimuthal variation of the power, temperature, and deformationof the assembly, highlighting behavior that is neglected in traditional axisymmetric fuelperformance codes that do not account for assembly features, such as guide tubesand control rods.11:10 AMA Coupled Neutronic/Thermal-Hydraulic Scheme BetweenCOBAYA3 and Subchanflow within the NURESIM SimulationPlatformManuel Calleja, Robert Stieglitz, Victor Sanchez, Javier Jimenez and UweImkeKarlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (INR)Multi-scale, multi-physics problems reveal significant challenges while dealing withcoupled neutronic/thermal-hydraulic solutions. Current generation of codes appliedto Light Water Reactors (LWR) are based on 3D neutronic nodal methods coupledwith one or two phase flow thermal-hydraulic system or sub-channel codes. In addition,spatial meshing and temporal schemes are crucial for the proper descriptionof the non-symmetrical core behavior in case of transient and accidents e.g. reactivityinsertion accidents. This paper describes the coupling approach between the 3Dneutron diffusion code COBAYA3 and the sub-channel code SUBCHANFLOW withinSALOME. The coupling is done inside the SALOME open source platform that is characterizedby a powerful pre- and post-processing capabilities and a novel functionalityfor mapping of the neutronic and thermal hydraulic domains. The peculiar functionalitiesof SALOME and the steps required for the code integration and coupling arepresented. The validation of the coupled codes is done based on two benchmarks thePWR MOX/UO2 RIA and the TMI-1 MSLB benchmark. A discussion of the predictioncapability of COBAYA3/SUBCHANFLOW compared to other coupled solutions will beprovided too.11:35 AMEvolution of Fast Reactor Core Spectra in Changing a HeavyLiquid Metal Coolant by Molten PB-208D.A. Blokhin, E.F. Mitenkova (1), G.L. Khorasanov, E.A. Zemskov, A.I.Blokhin (2)1) Nuclear Safety Institute of Russian Academy of Sciences, Moscow, Russia. 2) State Scientific Centerof the Russian Federation - Institute of Physics and Power, Obninsk, Kaluga Region, Russia10:45 AMPHISICS Toolkit: Multi-Reactor Transmutation Analysis Utility- MRTAUAndrea Alfonsi, Cristian Rabiti, Aaron S. Epiney, Yaqi Wang, Joshua CogliatiIdaho National Laboratory, Idaho Falls, ID, USAThe principal idea of this paper is to present the new capabilities available in thePHISICS toolkit, connected with the implementation of the depletion code MRTAU,a generic depletion/ decay/burn-up code developed at the Idaho National Laboratory.It is programmed in a modular structure and modern FORTRAN 95/2003. The codetracks the time evolution of the isotopic concentration of a given material accountingfor nuclear reactions happening in presence of neutron flux and also due to naturaldecay. MRTAU has two different methods to perform the depletion calculation, in orderto let the user choose the best on with respect to his needs. Both the methodologiesand some significant results are reported in this paper.In the paper neutron spectra of fast reactor cooled with lead–bismuth or lead-208 aregiven. It is shown that in changing the coolant from lead-bismuth to lead-208 the coreneutron spectra of the fast reactor FR RBEC-M are hardening in whole by severalpercents when a little share of low energy neutrons (5eV – 50 keV) is slightly increasing.The shift of spectra to higher energies permits to enhance the fuel fission whilethe increased share of low energy neutrons provides more effective conversion ofuranium-238 into plutonium due to peculiarity of 238U neutron capture cross section.Good neutron and physical features of molten 208Pb permit to assume it as perspectivecoolant for fast reactors and accelerator driven systems. The one-group crosssections of neutron radiation capture, , by 208Pb, 238U, 99Tc, mix of leadand bismuth, natPb-Bi, averaged over neutron spectra of the fast reactor RBEC-Mare given. It is shown that one-group cross sections of neutron capture by materialof the liquid metal coolant consisted from lead enriched with the stable lead isotope,208Pb, are by 4-7 times smaller than for the coolant natPb-Bi. The economyof neutrons in the core cooled with 208Pb can be used for reducing reactor’s initialfuel load, increasing fuel breeding and transmutation of long lived fission products,for example 99Tc. Good neutron and physical features of lead enriched with 208Pbpermit to consider it as a perspective low neutron absorbing coolant for fast reactorsand accelerator driven systems.83

Dr. Michael G. HoutsNuclear Research ManagerNASA Marshall Space Flight CenterPHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 -12:00 PM - Ballroom A-DThursday LuncheonPresentation: “Fission Power in Space”Introduction by Chris RobinsonDr. Michael G. Houts serves as Nuclear Research Manager at the NASA Marshall Space Flight Center, andis currently manager of NASA’s Nuclear Cryogenic Propulsion Stage (NCPS) project. Dr. Houts has over 20years of experience in the field of nuclear engineering, specializing in space nuclear power and propulsion.He received his PhD in Nuclear Engineering from the Massachusetts Institute of Technology. Dr. Houtsworked at Los Alamos National Laboratory for 11 years, where he held a variety of positions including Team Leader for Criticality, Reactor,and Radiation Physics and Deputy Group Leader of the 70 person Nuclear Design and Risk Analysis group. In these positions Dr. Houtswas involved in managing a variety of projects, many of them progressing from conceptual design to operational hardware. Dr. Houts hasbeen employed by NASA for 10 years. He is a member of the American Nuclear Society and the American Institute of Aeronautics andAstronautics. Dr. Houts has authored papers on a variety of subjects, including space nuclear power and propulsion, actinide transmutation,tritium production, radiation shielding, and others.The presentation will be followed byPHYSOR 2014 Informational SessionLed by ANS Reactor Physics Division representative.84

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 1:30 PM - 301 A1H - Core Analysis MethodsSession Chair: W. Zwermann (GRS); Ben Collins (U Mich)1:30 PMOn the Possible Gain of the Application of the SP3 SolutioninNodal CodesBruno MerkHelmholtz-Zentrum Dresden-Rossendorf, Institut für Sicherheitsforschung, Dresden, GermanyThe simplified PN method has attracted much attention in reactor physics. The implementationof an increased order expansion has the potential to improve the results.It is investigated if this potential can be released under the specific features of nodalcodes with fuel assembly sized calculation cells. One dimensional analytical solutionsfor the one and two energy-group diffusion, SP3 and P3 equations are derived. Theresults are compared to reference solutions calculated with ONEDANT and HELIOS.Tests are performed on the Brantley/Larsen one-group test case and on a pin cellwith real reactor materials. For the evaluation of the differences between the diffusionand the SP3 solutions in the nodal code configuration, benchmark configurations areinvestigated. A comparison of the diffusion and the SP3 results show improvementsfor the test case using fuel element size nodes, where the reference solution uses theidentical cross section basis. Comparing with a multi-group reference solution showsthat the diffusion solution sometimes behaves better due to error cancellation. Toachieve the full gain of the SP3 method in nodal calculations, a new standard for lightwater reactor calculations has to be defined with a refined energy-group structure.1:55 PMGPU-Accelerated 3D Neutron Diffusion Code Based on FiniteDifference MethodQi Xu, Ganglin Yu, Kan WangDepartment of Engineering Physics, Tsinghua UniversityFinite difference method, as a traditional numerical solution to neutron diffusion equation,although considered simpler and more precise than the coarse mesh nodalmethods, has a bottle neck to be widely applied caused by the huge memory andunendurable computation time it requires. In recent years, the concept of General-Purpose computation on GPUs has provided us with a powerful computational enginefor scientific research. In this study, a GPU-Accelerated multi-group 3D neutron diffusioncode based on finite difference method was developed. First, a clean-sheetneutron diffusion code (3DFD-CPU) was written in C++ on the CPU architecture, andlater ported to GPUs under NVIDIA’s CUDA platform (3DFD-GPU). The IAEA 3D PWRbenchmark problem was calculated in the numerical test, where three different codes,including the original CPU-based sequential code, the HYPRE (High PerformancePreconditioners)-based diffusion code and CITATION, were used as counterpoints totest the efficiency and accuracy of the GPU-based program. The results demonstrateboth high efficiency and adequate accuracy of the GPU implementation for neutrondiffusion equation. A speedup factor of about 46 times was obtained, using NVIDIA’sGeforce GTX470 GPU card against a 2.50 GHz Intel Quad Q9300 CPU processor.Compared with the HYPRE-based code performing in parallel on an 8-core towerserver, the speedup of about 2 still could be observed. More encouragingly, withoutany mathematical acceleration technology, the GPU implementation ran about 5 timesfaster than CITATION which was speeded up by using the SOR method and Chebyshevextrapolation technique.2:20 PMVerification of a Neutronic Code for Transient Analysis in ReactorsWith HEX-z GeometryS. González-Pintor and G. Verdú (1), D. Ginestar (2)1) Departamento de Ingeniería Química y Nuclear, Universitat Politècnica de València González, Valencia.Spain. 2) Departamento de Matemática Aplicada, Universitat Politècnica de València González,Valencia. SpainDue to the geometry of the fuel fundles, to simulate reactors such as VVER reactorsit is necessary to develop methods that can deal with hexagonal prisms as basic elementsof the spatial discretization. The main features of a code based on a high ordernite element method for the spatial discretization of the neutron diusion equation andan implicit dierence method for the time discretization of this equation are presentedand the performance of the code is tested solving the rst exercise of the AER transientbenchmark. The obtained results are compared with the reference results of thebenchmark and with the results provided by PARCS code.2:45 PMThe First-Principle Coupled Calculations Using TMCC andCFX for the Pin-Wise Simulation Of LWRLinsen Li and Kan WangDepartment of Engineering Physics, Tsinghua University, Department of Engineering Physics, Beijing,P.R. ChinaThe coupling of neutronics and thermal-hydraulics plays an important role in the reactorsafety, core design and operation of nuclear power facilities. This paper introducesthe research on the coupling of Monte Carlo method and CFD method, specificallyusing TMCC and CFX. The methods of the coupling including the coupling approach,data transfer, mesh mapping and transient coupling scheme are studied firstly. Thecoupling of TMCC and CFX for the steady state calculations is studied and describedfor the single rod model and the 3×3 Rod Bundle model. The calculation results provethat the coupling method is feasible and the coupled calculation can be used for steadystate calculations. However, the oscillation which occurs during the coupled calculationindicates that this method still needs to be improved for the accuracy. Then thecoupling for the transient calculations is also studied and tested by two cases of thesteady state and the lost of heat sink. The preliminary results of the transient coupledcalculations indicates that the transient coupling with TMCC and CFX is able to simulatethe transients but instabilities are occurring. It is also concluded that the transientcoupling of TMCC and CFX needs to be improved due to the limitation of computationalresource and the difference of time scales.3:10 PMA High-Fidelity Monte Carlo Evaluation of CANDU-6 SafetyParametersYonghee Kim and Donny HartantoKorea Advanced Institute of Science and Technology (KAIST), Daejeon, KoreaImportant safety parameters such as the fuel temperature coefficient (FTC) and thepower coefficient of reactivity (PCR) of the CANDU-6 (CANada Deuterium Uranium)reactor have been evaluated by using a modified MCNPX code. For accurate analysisof the parameters, the DBRC (Doppler Broadening Rejection Correction) schemewas implemented in MCNPX in order to account for the thermal motion of the heavyuranium nucleus in the neutron-U scattering reactions. In this work, a standard fuellattice has been modeled and the fuel is depleted by using the MCNPX and the FTCvalue is evaluated for several burnup points including the mid-burnup representing anear-equilibrium core. The Doppler effect has been evaluated by using several crosssection libraries such as ENDF/B-VI, ENDF/B-VII, JEFF, JENDLE. The PCR valueis also evaluated at mid-burnup conditions to characterize safety features of equilibriumCANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, hugenumber of neutron histories are considered in this work and the standard deviationof the k-inf values is only 0.5~1 pcm. It has been found that the FTC is significantlyenhanced by accounting for the Doppler broadening of scattering resonance and thePCR are clearly improved.85

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 1:30 PM - 301 B12D - Sensitivity & Uncertainty AnalysisSession Chair: Hany Abdel-Khalik (NCSU)1:30 PMUncertainty Analysis of a SFR Core with Sodium PlenumE. Canuti, E. Ivanov, V. Tiberi And S. PignetInstitut de Radioprotection et de Sûreté Nucléaire (IRSN), Fontenay-aux-Roses, FranceThe new concepts of Sodium-cooled Fast Reactors have to reach the Generation IVsafety objectives. In this regard the Sodium Void Effect has to be minimized for thefuture projects of large-size SFR as well as the uncertainties on it. The Institute ofRadiological Protection and Nuclear Safety (IRSN) as technological support of Frenchpublic authorities is in charge of safety assessment of operating and under constructionreactors, as well as future projects. In order to state about the safety of new SFRdesigns the IRSN must be able to evaluate core parameters and their uncertainties.In this frame a sensitivity and uncertainty study has been performed to evaluate theimpact of nuclear data uncertainty on sodium void effect, for the benchmark modelof large SFR BN-800. The benchmark parameters (effective multiplication factor andsodium void effect) have been evaluated using two codes, the deterministic code ERA-NOS and the Monte Carlo code SCALE, while the S/U analysis has been performedonly with SCALE. The results of the these studies point out the most relevant crosssection uncertainties that affect the SVE and how efforts should be done in increasingthe existing nuclear data accuracies.1:55 PMNuclear Data Uncertainties by the PWR MOX/UO2 Core RodEjection BenchmarkI. Pasichnyk, M. Klein, K. Velkov, W. Zwermann, A. PautzBoltzmannstr, Garching b. München, GERMANYRod ejection transient of the OECD/NEA and U.S. NRC PWR MOX/UO2 core benchmarkis considered under the influence of nuclear data uncertainties. Using the GRSuncertainty and sensitivity software package XSUSA the propagation of the uncertaintiesin nuclear data up to the transient calculations are considered. A statistically representativeset of transient calculations is analyzed and both integral as well as localoutput quantities are compared with the benchmark results of different participants. Itis shown that the uncertainties in nuclear data play a crucial role in the interpretationof the results of the simulation.2:20 PMNuclear Data Uncertainty Propagation in a Lattice PhysicsCode Using Stochastic SamplingW. Wieselquist, A. Vasiliev, and H. FerroukhiPaul Scherrer Institut, Villigen, SwitzerlandA methodology is presented for “black box” nuclear data uncertainty propagation in alattice physics code using stochastic sampling. The methodology has 4 components:i) processing nuclear data variance/covariance matrices including converting the nativegroup structure to a group structure “compatible” with the lattice physics code, ii)generating (relative) random samples of nuclear data, iii) perturbing the lattice physicscode nuclear data according to the random samples, and iv) analyzing the distributionof outputs to estimate the uncertainty. The scheme is described as implemented atPSI, in a modified version of the lattice physics code CASMO-5M, including all relevantpractical details. Uncertainty results are presented for a BWR pincell at hot zero powerconditions and a PWR assembly at hot full power conditions with depletion. Resultsare presented for uncertainties in eigenvalue, 1-group microscopic cross sections,2-group macroscopic cross sections, and isotopics. Interesting behavior is observedwith burnup, including a minimum uncertainty due to the presence of fertile U-238and a global effect described as “synergy”, observed when comparing the uncertaintyresulting from simultaneous and one-at-a-time variations of nuclear data.2:45 PMSecond Order Perturbation Theory for Nonlinear Time-DependentProblems: Application to a Simplified Coupled TransientGilli L., Lathouwers D., Kloosterman J.L. and T.H.J.J. van der HagenDelft University of Technology, Faculty of Applied Sciences Department of Radiation, Delft, The NetherlandsIn this paper a second-order perturbation technique for nonlinear time-dependentproblems is presented and applied to a simplified multi-physics model. This method isdeveloped by using the properties of the adjoint problem which allows calculating theset of first and second order coefficients by solving a number of linear systems. As anillustrative example the adjoint procedure is applied to a reference transient problem,represented by a coupled point-kinetic/lumped-parametersmodel, and used to calculatethe sensitivity coefficients of a safety related response with respect to a set ofinput parameters. The results obtained are compared with the values given by a directsampling of the forward nonlinear problem. A way to reduce the number of calculationsrequired for the application of second order adjoint techniques is also discussed. Ourfirst results show that the procedure provides good estimations in presence of higherorder perturbation components, being able to reconstruct the responses of interesteven in presence of non-Gaussian probability density functions. Furthermore, the useof reduced second order information decreases the computational requirements of themethod, making it appealing for possible large scale applications.86

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 1:30 PM - 301 C14C - Reactor Transient & Safety AnalysisSession Chair: Bill Williamson (TVA); Sara Bortot (KTH)1:30 PMSome Remarks to Application of Bifurcation Analysis In BWRStability AnalysisCarsten Lange, Dieter Hennig and Antonio HurtadoDepartment of Hydrogen and Nuclear Energy, Institute of Power Engineering, Technische UniversitätDresden, Germany2:45 PMAn Overview of the ENEA Activities in the Field of CoupledCodes NPP SimulationC. Parisi, E. Negrenti and M. Sepielli (1), A. Del Nevo (2)1) ENEA “Casaccia” Research Center, Rome, Italy. 2) ENEA “Brasimone” Research Center Camugnano,ItalyCurrently, BWR stability analysis is most often performed by application of systemcodes which provide the time evolution of the neutron flux or thermal power at a definedoperational point (OP) after imposing a system parameter perturbation. However,in general, it is impossible to understand the real stability state of the BWR at a specificOP by application of system code analysis alone. Hence, we are searching for methodsdeveloped in the nonlinear dynamics field in order to reveal the nature of the BWRstability states when power oscillations are observed. A powerful method is bifurcationanalysis. In this paper, we will demonstrate an example of a possible BWR instabilityphenomenon, which can be understood only in nonlinear terms.1:55 PMInterpretation of Experimental Data on Safety Parametersfrom Standpoint of Mathematical Nuclear Reactor TheoryAlexey L. Cherezov, Nikolai V. ShchukinNational Nuclear Research University “MEPhI”, Kashirskoe, RussiaThe paper deals with the problem of correct processing of experimental informationfor unambiguous determination of current reactivity value. The methodology for reactivitydetermination that is based on spectral decomposition of the solution obtainedfor non-stationary equation of neutron transport is presented. The proposed spectralprojection algorithm takes into consideration the effects related with space-time redistributionof neutron field and thus excludes the main systematic error and increasesaccuracy of current reactivity values. The paper presents the results of model calculations,demonstrating the effectiveness of the proposed method.2:20 PMIdentification of Limiting Case Between DBA and SBDBA (CLBreak Area Sensitivity): A New Model for the Boron InjectionSystemR. Gonzalez Gonzalez, A. Petruzzi and F. D’Auria (1), O. Mazzantini (2)1) San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Pisa, ITALY. 2) NucleoelectricaArgentina Sociedad Anonima (NA-SA), Buenos Aires, ARGENTINAIn the framework of the nuclear research activities in the fields of safety, training andeducation, ENEA (the Italian National Agency for New Technologies, Energy and theSustainable Development) is in charge of defining and pursuing all the necessarysteps for the development of a NPP engineering simulator at the “Casaccia” ResearchCenter near Rome. A summary of the activities in the field of the nuclear power plantssimulation by coupled codes is here presented with the long term strategy for theengineering simulator development. Specifically, results from the participation in internationalbenchmarking activities like the OECD/NEA “Kalinin-3” benchmark and the“AER-DYN-002” benchmark, together with simulations of relevant events like the Fukushimaaccident, are here reported. The ultimate goal of such activities performed usingstate-of-the-art technology is the re-establishment of top level competences in theNPP simulation field in order to facilitate the development of Enhanced EngineeringSimulators and to upgrade competences for supporting national energy strategy decisions,the nuclear national safety authority, and the R&D activities on NPP designs.3:10 PMAlgorithm Development And Verification Of UASCM for Multi-Diemnsion and Multi-Group Neutron Kinetics ModelShengyi SihShanghai Nuclear Engineering Research and Design Institute, Shanghai, ChinaThe Universal Algorithm of Stiffness Confinement Method (UASCM) for neutron kineticsmodel of multi-dimensional and multi-group transport equations or diffusion equationshas been developed. The numerical experiments based on transport theory codeMGSNM and diffusion theory code MGNEM have demonstrated that the algorithm hassufficient accuracy and stability.Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic ofArgentina. Its geometrical complexity and (e.g., oblique Control Rods, Positive Void coefficient)required a developed and validated complex three dimensional (3D) neutronkinetics (NK) coupled thermal hydraulic (TH) model. Reactor shut-down is obtained byoblique CRs and, during accidental conditions, by an emergency shut-down system(JDJ) injecting a highly concentrated boron solution (boron clouds) in the moderatortank, the boron clouds reconstruction is obtained using a CFD (CFX) code calculation.A complete LBLOCA calculation implies the application of the RELAP5-3D© systemcode. Within the framework of the third Agreement “NA-SA – University of Pisa” a newRELAP5-3D control system for the boron injection system was developed and implementedin the validated coupled RELAP5-3D/NESTLE model of the Atucha 2 NPP.The aim of this activity is to find out the limiting case (maximum break area size) for thePeak Cladding Temperature for LOCAs under fixed boundary conditions.87

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 1:30 PM - 301 D15C - Experimental Facilities & ExperimentsSession Chair: Frederik Reitsma (Calvera Consultants); M. Carta (ENEA)1:30 PMReactivity Loss Validation Of High Burn-Up PPWR Fuels withPile-Oscillation Experiments in MinervePierre Leconte, Claire Vaglio-Gaudard, Romain Eschbach, Jacques Di-Salvo, Muriel Antony, Alexandra PepinoCEA, DEN, DER, Cadarache, F-13108 Saint-Paul-Lez-DuranceThe ALIX experimental program relies on the experimental validation of the spentfuel inventory, by chemical analysis of samples irradiated in a PWR between 5 and7 cycles, and also on the experimental validation of the spent fuel reactivity loss withburn-up, obtained by pile-oscillation measurements in the MINERVE reactor. Theselatter experiments provide an overall validation of both the fuel inventory and of thenuclear data responsible for the reactivity loss. This program offers also unique experimentaldata for fuels with a burn-up reaching 85 GWd/t, as spent fuels in FrenchPWRs never exceeds 70 GWd/t up to now. The analysis of these experiments is donein two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first one, thefuel inventory of each sample is obtained by assembly calculations. The calculationroute consists in the self-shielding of cross sections on the 281 energy group SHEMmesh, followed by the flux calculation by the Method Of Characteristics in a 2D-exactheterogeneous geometry of the assembly, and finally a depletion calculation by aniterative resolution of the Bateman equations. In the second step, the fuel inventory isused in the analysis of pile-oscillation experiments in which the reactivity of the ALIXspent fuel samples is compared to the reactivity of fresh fuel samples. The comparisonbetween Experiment and Calculation shows satisfactory results with the JEFF3.1.1library which predicts the reactivity loss within 2% for burn-up of ~75 GWd/t and within4% for burn-up of ~85 GWd/t.1:55 PMTheoretical Analysis of the Subcritical Experiments Performedin the IPEN/MB-01 Research Reactor FacilitySeung Min Lee and Adimir dos SantosInstituto de Pesquisas Energéticas e Nucleares, São Paulo, BRASILThe theoretical analysis of the subcritical experiments performed at the IPEN/MB-01reactor employing the coupled NJOY/AMPX-II/TORT systems was successfully accomplished.All the analysis was performed employing ENDF/B-VII.0. The theoreticalapproach follows all the steps of the subcritical model of Gandini and Salvatores. Thetheory/experiment comparison reveals that the calculated subcritical reactivity is ina very good agreement to the experimental values. The subcritical index ( ) showssome discrepancies although in this particular case some work still have to be madeto model in a better way the neutron source present in the experiments.2:20 PMReactivity Worth Measurements on Fast Burst Reactor Caliban- Description and Interpretation of Integral Experimentsfor the Validation of Nuclear DataBenoît RichardCommissariat à l’Energie Atomiqueet aux Energies Alternatives, CEA, DAM, VALDUC, Is-sur-Tille,France.Reactivity perturbation experiments using various materials are being performed onthe HEU fast core CALIBAN, an experimental device operated by the CEA VALDUCCriticality and Neutron Transport Research Laboratory. These experiments providevaluable information to contribute to the validation of nuclear data for the materialsused in such measurements. This paper presents the results obtained in a first seriesof measurements performed with Au-197 samples. Experiments which have beenconducted in order to improve the characterization of the core are also described anddiscussed. The experimental results have been compared to numerical calculation usingboth deterministic and Monte Carlo neutron transport codes with a simplified modelof the reactor. This early work led to a methodology which will be applied to the futureexperiments which will concern other materials of interest.2:45 PMAnalysis of PIN Removal Experiments Conducted in an SCWR-Like Test LatticeR. Chawla, D. Rätz, K. A. Jordan, G. PerretPaul Scherrer Institue, Villigen PSI, SwitzerlandA comprehensive program of integral experiments, largely based on the measurementof reaction rate distributions, was carried out recently on an SCWR-like fuel lattice inthe central test zone of the PROTEUS zero-power research reactor at the Paul ScherrerInstitute in Switzerland. The present paper reports on the analysis of a complementaryset of measurements, in which the reactivity effects of removing individual pinsfrom the unperturbed, heterogeneously moderated reference lattice were investigated.It has been found that the detailed Monte Carlo modeling of the whole reactor usingMCNPX is able – as in the case of the reaction rate distributions – to reproduce the experimentalresults for the pin removal worths within the achievable statistical accuracy.A comparison of reducedgeometry calculations between MCNPX and the deterministicLWR assembly code CASMO-4E has revealed certain discrepancies. On the basis ofa reactivity decomposition analysis of the differences between the codes, it has beensuggested that these could be due to CASMO-4E deficiencies in calculating the effect,upon pin removal, of the extra moderation in the neighboring fuel pins.3:10 PMThe Second and Third NGNP Advanced Gas Reactor Fuel IrradiationExperimentsS. B. Grover and D. A. PettiIdaho National Laboratory, Idaho Falls, IDThe United States Department of Energy’s Next Generation Nuclear Plant (NGNP) AdvancedGas Reactor (AGR) Fuel Development and Qualification Program is currentlyscheduled to irradiate a total of five low enriched uranium (LEU) tri-isotopic (TRISO)particle fuel experiments in the Advanced Test Reactor (ATR) located at the Idaho NationalLaboratory (INL). The irradiations are being accomplished to demonstrate andqualify new TRISO coated particle fuel for use in high temperature gas cooled reactors.The experiments will each consist of at least six separate capsules, and will be irradiatedin an inert sweep gas atmosphere with individual on-line temperature monitoringand control of each capsule. The effluent sweep gas will also have on-line fissionproduct monitoring to track performance of the fuel in each individual capsule duringirradiation. The first experiment (designated AGR-1) started irradiation in December2006 and completed a very successful irradiation in early November 2009. The secondexperiment (AGR-2) started irradiation in June 2010, and the third and fourth experimentshave been combined into a single larger irradiation (AGR-3/4) that is currentlybeing assembled. The design and status of the second through fourth experiments aswell as the irradiation results of the second experiment to date are discussed.88

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 1:30 PM - 301 E8G - Advanced Modeling & Simulation in Reactor PhysicsSession Chair: Tom Sutton (KAPL)1:30 PMBoron Injection/Dilution Capabilities in TRACB/NEM CoupledCodeA. Jambrina, T. Barrachina, R. Miró, G. VerdúInstitute for the Industrial, Radiophysical and Environmental Safety (ISIRYM), Universitat Politècnica deValència (UPV)The coupled code TRAC-BF1/NEM is a thermal-hydraulic-neutronic code which allowstransient simulations considering neutronic 3D and thermal-hydraulic process in multiplechannels with one-dimensional geometry. TRAC-BF1 and NEM can be executedeither in stand-alone mode, i.e. without coupling, as well as coupled. In stand-alonecalculations NEM code is used without coupling and the thermal-hydraulic conditions(fuel temperature, moderator density and boron concentration) and xenon concentrationfor each node are taken from the SIMULATE3 output files. The NEM’s source codehas been modified to be able to read these conditions from external files when it isexecuted without being coupled. The coupling between TRAC-BF1 and NEM followsan integration scheme in which the thermal-hydraulic solution of TRAC-BF1 is sentto NEM to incorporate the feedback effects through the cross sections. TRAC-BF1solves heat conduction equations inside of the heat structures using the 3D powerdistribution from NEM. The coupling is carried out through the communication protocolfunctions of PVM (Parallel Virtual Machine). The present article presents a study whichconstitutes an advance in the simulation of injection, transport and mix of boron inthe reactor, increasing the capabilities of TRAC-BF1/NEM coupled code. This articleshows the modifications introduced in the TRAC-BF1/NEM’s source code to allow amore realistic simulation of boron injection transients. The qualification of these improvementsin both codes is performed simulating a steady state of a generic BWR atnominal power. The results have been compared with SIMULATE3 which is used as areference to obtain the cross sections through the SIMTAB methodology.1:55 PMDevelopment of a New Methodology for Stability Analysis inBWR NPPM. Garcia-Fenoll, A. Abarca, T. Barrachina, R. Miró, G. VerdúInstitute for Industrial, Radiophysical and Environmental Safety (ISIRYM), Universitat Politècnica deValència, València, SpainIn this work, a new methodology to reproduce power oscillations in BWR NPP ispresented. This methodology comprises the modal analysis techniques, the signalanalysis techniques and the simulation with the coupled code RELAP5/PARCSv2.7.Macroscopic cross sections are obtained by using the SIMTAB methodology, whichis fed up with CASMO-4/SIMULATE-3 data. The input files for the neutronic and thermohydrauliccodes are obtained automatically and the thermalhydraulic-to-neutronicrepresentation (mapping) used is based on the fundamental, first and second harmonicsshapes of the reactor power, calculated with the VALKIN code (developed in UPV).This mapping was chosen in order not to condition the oscillation pattern. To introducepower oscillations in the simulation a new capability in the coupled code, for generatedensity perturbations (both for the whole core and for chosen axial levels) accordingwith the power modes shapes, has been implemented. The purpose of the methodologyis to reproduce the driving mechanism of the out of phase oscillations appearedin BWR type reactors. In this work, the methodology is applied to the Record 9 point,collected in the NEA benchmark of Ringhals 1 NPP. A set of different perturbations areinduced in the first active axial level and the LPRM signals resulting are analyzed.2:45 PMEigenvalue Sensitivity Studies for the Fort St. Vrain High TemperatureGas-Cooled Reactor to Account for Fabrication andModeling UncertaintiesAndrew T. Pavlou, Benjamin R. Betzler, Timothy P. Burke, John C. Lee, WilliamR. Martin, Wilson N. Pappo, and Eva E. SunnyUniversity of Michigan, Department of Nuclear Engineering & Radiological Sciences, Ann Arbor, MIUncertainties in the composition and fabrication of fuel compacts for the Fort St. Vrain(FSV) high temperature gas reactor have been studied by performing eigenvalue sensitivitystudies that represent the key uncertainties for the FSV neutronic analysis. Theuncertainties for the TRISO fuel kernels were addressed by developing a suite of modelsfor an “average” FSV fuel compact that models the fuel as (1) a mixture of two differentTRISO fuel particles representing fissile and fertile kernels, (2) a mixture of fourdifferent TRISO fuel particles representing small and large fissile kernels and smalland large fertile kernels and (3) a stochastic mixture of the four types of fuel particleswhere every kernel has its diameter sampled from a continuous probability densityfunction. All of the discrete diameter and continuous diameter fuel models were constrainedto have the same fuel loadings and packing fractions. For the non-stochasticdiscrete diameter cases, the MCNP compact model arranged the TRISO fuel particleson a hexagonal honeycomb lattice. This lattice-based fuel compact was compared toa stochastic compact where the locations (and kernel diameters for the continuous diametercases) of the fuel particles were randomly sampled. Partial core configurationswere modeled by stacking compacts into fuel columns containing graphite. The differencesin eigenvalues between the lattice-based and stochastic models were small butthe runtime of the lattice-based fuel model was roughly 20 times shorter than with thestochastic-based fuel model.3:10 PMSimulation of In-Core Neutron Noise Measurements for AxialVoid Profile Reconstruction in Boiling Water ReactorsV. Dykin and I. PázsitChalmers University of Technology, Division of Nuclear Engineering, Gothenburg, SwedenA possibility to reconstruct the axial void profile from the simulated in-core neutronnoise which is caused by density fluctuations in a Boiling Water Reactor (BWR) heatedchannel is considered. For this purpose, a self-contained model of the two-phase flowregime is constructed which has quantitatively and qualitatively similar properties tothose observed in real BWRs. The model is subsequently used to simulate the signalsof neutron detectors induced by the corresponding perturbations in the flow density.The bubbles are generated randomly in both space and time using Monte-Carlo techniques.The axial distribution of the bubble production is chosen such that the meanaxial void fraction and void velocity follow the actual values of BWRs. The inducedneutron noise signals are calculated and then processed by the standard signal analysismethods such as Auto-Power Spectral Density (APSD) and Cross-Power SpectralDensity (CPSD). Two methods for axial void and velocity profiles reconstruction arediscussed: the first one is based on the change of the break frequency of the neutronauto-power spectrum with axial core elevation, while the second refers to the estimationof transit times of propagating steam fluctuations between different axial detectorpositions. This paper summarizes the principles of the model and presents a numericaltesting of the qualitative applicability to estimate the required parameters for the reconstructionof the void fraction profile from the neutron noise measurements.2:20 PMA Newton-Krylov Solution to the Porous Medium Equations inthe AGREE CodeAndrew M. Ward, Volkan Seker, Yunlin Xu, and Thomas J. DownarDepartment of Nuclear Engineering and Radioligical Sciences, University of Michigan, Ann Arbor, MIIn order to improve the convergence of the AGREE code for porous medium, a Newton-Krylovsolver was developed for steady state problems. The current three-equationsystem was expanded and then coupled using Newton’s Method. Theoretical behaviorpredicts second order convergence, while actual behavior was highly nonlinear. Thediscontinuous derivatives found in both closure and empirical relationships preventedtrue second order convergence. Agreement between the current solution and new ExactNewton solution was well below the convergence criteria. While convergence timedid not dramatically decrease, the required number of outer iterations was reduced byapproximately an order of magnitude. GMRES was also used to solve problem, whereILU without fill-in was used to precondition the iterative solver, and the performancewas slightly slower than the direct solution.89

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 3:50 PM - 301 A10A - Nuclear Criticality SafetySession Chair: M. Brady-Rapp (PNNL); Brad Rearden (ORNL)3:50 PMBurst Wait Time Simulation of Caliban Reactor at DelayedSuper-Critical StatePhilippe Humbert (1), Nicolas Authier, Benoît Richard, Pascal Grivot, PierreCasoli (2)1) Commissariat à l’Energie Atomique, Arpajon, France. 2) Commissariat à l’Energie Atomique, Is-sur-Tille, FranceIn the past, the super prompt critical wait time probability distribution was measured onCALIBAN fast burst reactor [4]. Afterwards, these experiments were simulated with avery good agreement by solving the non-extinction probability equation [5]. Recently,the burst wait time probability distribution has been measured at CEA-Valduc on CALI-BAN at different delayed super-critical states [6]. However, in the delayed super-criticalcase the non-extinction probability does not give access to the wait time distribution.In this case it is necessary to compute the time dependent evolution of the full neutroncount number probability distribution. In this paper we present the point model deterministicmethod used to calculate the probability distribution of the wait time beforea prescribed count level taking into account prompt neutrons and delayed neutronprecursors. This method is based on the solution of the time dependent adjoint Kolmogorovmaster equations for the number of detections using the generating functionmethodology [8,9,10] and inverse discrete Fourier transforms. The obtained resultsare then compared to the measurements and Monte-Carlo calculations based on thealgorithm presented in [7].4:15 PMHTTR Criticality Calculations with SCALE6: Studies of VariousGeometric and Unit-Cell Options in ModelingJui-Yu Wang, Min-Han Chiang, Rong-Jiun Sheu, Yen-Wan Hsueh LiuInstitute of Nuclear Engineering and Science, National Tsing Hua University, Hsinchu, TaiwanThe fuel element of the High Temperature Engineering Test Reactor (HTTR) presentsa doubly heterogeneous geometry, where tiny TRISO fuel particles dispersed in agraphite matrix form the fuel region of a cylindrical fuel rod, and a number of fuel rodstogether with moderator or reflector then constitute the lattice design of the core. Inthis study, a series of full-core HTTR criticality calculations were performed with theSCALE6 code system using various geometric and unit-cell options in order to systematicallyinvestigate their effects on neutronic analysis. Two geometric descriptions(ARRAY or HOLE) in SCALE6 can be used to construct a complicated and repeatedmodel. The result shows that eliminating the use of HOLE in the HTTR geometricmodel can save the computation time by a factor of 4. Four unit-cell treatments for resonanceself-shielding corrections in SCALE6 were tested to create problem-specificmultigroup cross sections for the HTTR core model. Based on the same ENDF/B-VIIcross-section library, their results were evaluated by comparing with continuous-energycalculations. The comparison indicates that the INFHOMMEDIUM result overestimatesthe system multiplication factor (keff) by 55 mk, whereas the LATTICECELL andMULTIREGION treatments predict the keff values with similar biases of approximately10 mk overestimation. The DOUBLEHET result shows a more satisfactory agreement,about 4.2 mk underestimation in the keff value. In addition, using cell-weighted crosssections instead of an explicit modeling of TRISO particles in fuel region can furtherreduce the computation time by a factor of 5 without sacrificing accuracy.5:05 PMNeutron Initiation Probability In Fast Burst ReactorLiu Xiaobo, Du Jinfeng, Xie Qilin, Fan XiaoqiangInstitute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Sichuan, P.R.ChinaBased on the probability balance of neutron random events in multiply system, the fourrandom process of neutron in prompt super-critical is described and then the equationof neutron initiation probability is deduced. On the assumption of static, slightly promptsuper-critical athe two factorial approximation, the formula of the average probabilityof “one” neutron is derivwhich is the same with the result derived from the pointmodel. The MC simulation using point model is applied in Godiva- and CFBR-, and thesimulation result of one neutron initiatiwell consistent with the theory that the initiationprobability of Godiva- CFBR-II burst reactorare 0.00032, 0.00027 respectively on theordinary burst operation.5:30 PMBenefits of the Delta K of Depletion Benchmarks for BurnupCredit ValidationDale Lancaster (1), Albert Machiels (2)1) NuclearConsultants.com, Boalsburg, PA. 2) Electric Power Research Institute, Inc., Palo Alto, CAPressurized Water Reactor (PWR) burnup credit validation is demonstrated using thebenchmarks for quantifying fuel reactivity decrements, published as “Benchmarks forQuantifying Fuel Reactivity Depletion Uncertainty,” EPRI Report 1022909 (August2011). This demonstration uses the depletion module TRITON available in the SCALE6.1 code system followed by criticality calculations using KENO-Va. The differencebetween the predicted depletion reactivity and the benchmark’s depletion reactivity isa bias for the criticality calculations. The uncertainty in the benchmarks is the depletionreactivity uncertainty. This depletion bias and uncertainty is used with the bias anduncertainty from fresh UO2 critical experiments to determine the criticality safety limitson the neutron multiplication factor, keff. The analysis shows that SCALE 6.1 with theENDF/B-VII 238-group cross section library supports the use of a depletion bias of only0.0015 in delta k if cooling is ignored and 0.0025 if cooling is credited. The uncertaintyin the depletion bias is 0.0064. Reliance on the ENDF/B V cross section library producesmuch larger disagreement with the benchmarks. The analysis covers numerouscombinations of depletion and criticality options. In all cases, the historical uncertaintyof 5% of the delta k of depletion (“Kopp memo”) was shown to be conservative for fuelwith more than 30 GWD/MTU burnup. Since this historically assumed burnup uncertaintyis not a function of burnup, the Kopp memo’s recommended bias and uncertaintymay be exceeded at low burnups, but its absolute magnitude is small.4:40 PMPWR ENDF/B-VII Cross-Section Libraries for Origen-ArpCarolyn McGraw (1), Germina Ilas (2)1) Department of Nuclear Engineering, Texas A&M University, College Station, TX. 2) Oak Ridge NationalLaboratory, Oak Ridge, TNNew pressurized water reactor (PWR) cross-section libraries were generated for usewith the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system.These libraries are based on ENDF/B-VII nuclear data and were generated usingthe two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The librariescontain multiple burnup-dependent cross-sections for seven PWR fuel designs, withenrichments ranging from 1.5 to 6 wt% 235U. The burnup range has been extendedfrom the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU.Validation of the libraries using radiochemical assay measurements and decay heatmeasurements for PWR spent fuel showed good agreement between calculated andexperimental data. Verification against detailed TRITON simulations for the consideredassembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with directTRITON depletion, while greatly reducing the computation time.90

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 3:50 PM - 301 B12E - Sensitivity & Uncertainty AnalysisSession Chair: Tatiana Ivanova (IRSN); Tom Sutton (KAPL)3:50 PMValidation of the U.S. NRC Coupled Code System TRITON/TRACE/PARCS with the Special Power Excursion ReactorTest III (SPERT III)Raymond C. Wang, Yunlin Xu, and Thomas Downar (1), Nathanael Hudson(2)1) Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI.2) RES Division, U.S. NRC, Rockville, MDThe Special Power Excursion Reactor Test III (SPERT III) was a series of reactivityinsertion experiments conducted in the 1950s. This paper describes the validation ofthe U.S. NRC Coupled Code system TRITON/PARCS/TRACE to simulate reactivityinsertion accidents (RIA) by using several of the SPERT III tests. The work here usedthe SPERT III E-core configuration tests in which the RIA was initiated by ejectinga control rod. The resulting superprompt reactivity excursion and negative reactivityfeedback produced the familiar bell shaped power increase and decrease. The energydeposition during such a power peak has important safety consequences and providesvalidation basis for core coupled multi-physics codes. The transients of five separatetests are used to benchmark the PARCS/TRACE coupled code. The models werethoroughly validated using the original experiment documentation.4:15 PMCASMO5/TSUNAMI-3D Spent Nuclear Fuel Reactivity UncertaintyAnalysisRodolfo Ferrer and Joel Rhodes (1), Kord Smith (2)Studsvik Scandpower, Inc., Idaho Falls, ID. 2) Department of Nuclear Science and Engineering, MassachusettsInstitute of Technology, Cambridge, MAThe CASMO5 lattice physics code is used in conjunction with the TSUNAMI-3D sequencein ORNL’s SCALE 6 code system to estimate the uncertainties in hot-to-coldreactivity changes due to cross-section uncertainty for PWR assemblies at variousburnup points. The goal of the analysis is to establish the multiplication factor uncertaintysimilarity between various fuel assemblies at different conditions in a quantifiablemanner and to obtain a bound on the hot-tocold reactivity uncertainty over the variousassembly types and burnup attributed to fundamental cross-section data uncertainty.4:40 PMUncertainty Quantification for Accident Management UsingAce SurrogatesAthi Varuttamaseni and John C. Lee (1), Robert W. Youngblood (2)Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI. 2)Idaho National Laboratory, Idaho Falls, IDThe alternating conditional expectation (ACE) regression method is used to generateRELAP5 surrogates which are then used to determine the distribution of the peak cladtemperature (PCT) during the loss of feedwater accident coupled with a subsequentinitiation of the feed and bleed (F&B) operation in the Zion-1 nuclear power plant. Theconstruction of the surrogates assumes conditional independence relations amongkey reactor parameters. The choice of parameters to model is based on the macroscopicbalance statements governing the behavior of the reactor. The peak cladtemperature is calculated based on the independent variables that are known to beimportant in determining the success of the F&B operation. The relationship betweenthese independent variables and the plant parameters such as coolant pressure andtemperature is represented by surrogates that are constructed based on 45 RELAP5cases. The time-dependent PCT for different values of F&B parameters is calculatedby sampling the independent variables from their probability distributions and propagatingthe information through two layers of surrogates. The results of our analysisshow that the ACE surrogates are able to satisfactorily reproduce the behavior of theplant parameters even though a quasi-static assumption is primarily used in their construction.The PCT is found to be lower in cases where the F&B operation is initiated,compared to the case without F&B, regardless of the F&B parameters used.5:05 PMSensitivity Analysis of Coupled Criticality CalculationsZoltán Perkó, Jan Leen Kloosterman and Danny LathouwersDelft University of Technology, Faculty of Applied Physics, Radionuclides and Reactors, Delft, The NetherlandsPerturbation theory based sensitivity analysis is a vital part of todays’ nuclear reactordesign. This paper presents an extension of standard techniques to examinecoupled criticality problems with mutual feedback between neutronics and an augmentingsystem (for example thermal-hydraulics). The proposed procedure uses aneutronic and an augmenting adjoint function to efficiently calculate the first orderchange in responses of interest due to variations of the parameters describing thecoupled problem. The effect of the perturbations is considered in two different waysin our study: either a change is allowed in the power level while maintaining criticality(power perturbation) or a change is allowed in the eigenvalue while the power isconstrained (eigenvalue perturbation). The calculated response can be the change inthe power level, the reactivity worth of the perturbation, or the change in any functionalof the flux, the augmenting dependent variables and the input parameters. To obtainpower- and criticality-constrained sensitivities power- and k-reset procedures can beapplied yielding identical results. Both the theoretical background and an applicationto a one dimensional slab problem are presented, along with an iterative procedureto compute the necessary adjoint functions using the neutronics and the augmentingcodes separately, thus eliminating the need of developing new programs to solve thecoupled adjoint problem.5:30 PMInteraction of Loading Pattern and Nuclear Data Uncertaintiesin Reactor Core CalculationsM. Klein, L. Gallner, B. Krzykacz-Hausmann, A. Pautz, K. Velkov, W. ZwermannGesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH Boltzmannstr. 14, Garching b. München,GERMANYAlong with best-estimate calculations for design and safety analysis, understandinguncertainties is important to determine appropriate design margins. In this framework,nuclear data uncertainties and their propagation to full core calculations are a criticalissue. To deal with this task, different error propagation techniques, deterministic andstochastic are currently developed to evaluate the uncertainties in the output quantities.Among these is the sampling based uncertainty and sensitivity software XSUSAwhich is able to quantify the influence of nuclear data covariance on reactor corecalculations. In the present work, this software is used to investigate systematicallythe uncertainties in the power distributions of two PWR core loadings specified inthe OECD UAM-Benchmark suite. With help of a statistical sensitivity analysis, themain contributors to the uncertainty are determined. Using this information a methodis studied with which loading patterns of reactor cores can be optimized with regardto minimizing power distribution uncertainties. It is shown that this technique is able tohalve the calculation uncertainties of a MOX/UOX core configuration.91

Session Chair: D. Cokinos (BNL)PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 3:50 PM - 301 C11D - Nuclear Data3:50 PMBenchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1Steven C. van der MarckNuclear Research & Consultancy Group (NRG), Petten, the NetherlandsThree nuclear data libraries have been tested extensively using criticality safetybenchmark calculations. The three libraries are the new release of the US libraryENDF/B-VII.1 (2011), the new release of the Japanese library JENDL-4.0 (2011),and the OECD/NEA library JEFF-3.1 (2006). All calculations were performed with thecontinuous-energy Monte Carlo code MCNP (version 4C3, as well as version 6-beta1).Around 2000 benchmark cases from the International Handbook of Criticality SafetyBenchmark Experiments (ICSBEP) were used. The results were analyzed per ICS-BEP category, and per element. Overall, the three libraries show similar performanceon most criticality safety benchmarks. The largest differences are probably caused byelements such as Be, C, Fe, Zr, W.4:15 PM183W Resonance Parameter Evaluation in the Neutron EnergyRange up to 5 keVM. T. Pigni, M. E. Dunn, K. H. GuberOak Ridge National Laboratory, Nuclear Data & Criticality Safety, Oak Ridge, TNWe generated a preliminary set of resonance parameters for 183W in the neutron energyrange of thermal up to 5 keV. In the analyzed energy range, this work representsa significant improvement over the current resonance evaluation in the ENDF/B-VII.1library limited up to 2.2 keV. The evaluation methodology uses the Reich-Moore approximationto fit, with the R-matrix code SAMMY, the high-resolution measurementsperformed in 2007 at the GEel LINear Accelerator (GELINA) facility. The transmissiondata and the capture cross sections calculated with the set of resonance parametersare compared with the experimental values, and the average properties of the resonanceparameters are discussed.4:40 PMConvergence of Legendre Expansion of Doppler-BroadenedDouble-Differential Elastic Scattering Cross SectionG. Arbanas, M. E. Dunn, N. M. Larson, L. C. Leal, and M. L. Williams (1), B.Becker (2), R. Dagan (3)1) Oak Ridge National Laboratory, Oak Ridge, TN. 2) Department of Mechanical, Aerospace and NuclearEngineering, Rensselaer Polytechnic Institute NES, Troy, NY. 3) Institut für Neutronenphysik und Reaktortechnik,Forschungszentrum Karlsruhe Gmbh, Karlsruhe, Germany5:05 PMEvaluating and Adjusting 239Pu, 56Fe, 28Si and 95Mo NuclearData with a Monte Carlo TechniqueD. Rochman and A.J. KoningNuclear Research and Consultancy Group, Petten, The NetherlandsIn this paper,Monte Carlo optimization and nuclear data evaluation are combined toproduce optimal adjusted nuclear data files. The methodology is based on the socalled”Total Monte Carlo” and the TALYS system. Not only a single nuclear data file isproduced for a given isotope, but virtually an infinite number, defining probability distributionsfor each nuclear quantity. Then each of these random nuclear data librariesis used in a series of benchmark calculations. With a goodness-of-fit estimator, best239Pu, 56Fe, 28Si and 95Mo evaluations for that benchmark set can be selected. Afew thousands of random files are used and each of them is tested with a large numberof fast, thermal and intermediate energy criticality benchmarks. From this, the bestperforming random file is chosen and proposed as the optimum choice among thestudied random set.5:30 PMProposal for the Utilization of the Total Cross Section Covariancesand its Correlations with Channel Reactions for Sensitivityand Uncertainty AnalysisP. Sabouri, A. Bidaud (1), S. Dabiran, A. Buijs (2)1) Labratoire de Physique Subatomique et de Cosmologie, CNRS-IN2P3/UJF/INPG, Grenoble, France.2) Department of Engineering Physics, McMaster University, Hamilton, CanadaAn alternate method for the estimation of the global uncertainty on criticality, using thetotal cross section and its covariances, is proposed. Application of the method withcurrently available covariance data leads to an unrealistically large prediction of theglobal uncertainty on criticality. New covariances for total cross section and individualreactions are proposed. Analysis with the proposed covariance matrices is found toresult in a global uncertainty for criticality consistent with the traditional method. Recommendationsare made to evaluators for providing total cross section covariances.Convergence properties of Legendre expansion of a Doppler-broadened double-differentialelastic neutron scattering cross section of 238U near the 6:67 eV resonanceat temperature 103 K are studied. A variance of Legendre expansion from a referenceMonte Carlo computation is used as a measure of convergence and is computed foras many as 15 terms in the Legendre expansion. When the outgoing energy equalsthe incoming energy, it is found that the Legendre expansion converges very slowly.Therefore, a supplementary method of computing many higher-order terms is suggestedand employed for this special case.92

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 3:50 PM - 301 D14D - Reactor Transient & Safety AnalysisSession Chair: Sara Bortot (KTH)3:50 PMTransition to CASMO-5M and SIMULATE-3K for Stability Analysesof the Swiss BWRsAbdelhamid Dokhane, Stefano Canepa and Hakim FerroukhiPaul Scherrer Institute, Villigen-PSI, SwitzerlandFor stability analyses of the Swiss operating Boiling-Water-Reactors (BWRs), themethodology employed and validated so far at the Paul Scherrer Institute (PSI) wasbased on the RAMONA-3 code with a hybrid upstream static lattice/core analysis approachusing CASMO-4 and PRESTO-2. More recently, steps were undertaken towardsa new methodology based on the SIMULATE-3K (S3K) code for the dynamicalanalyses combined with the CMSYS system relying on the CASMO/SIMULATE-3suite of codes and which was established at PSI to serve as framework for the developmentand validation of reference core models of all the Swiss reactors and operatedcycles. This paper presents a first validation of the new methodology on the basisof a benchmark recently organised by a Swiss utility and including the participationof several international organisations with various codes/methods. Now in parallel, atransition from CASMO-4E (C4E) to CASMO-5M (C5M) as basis for the CMSYS coremodels was also recently initiated at PSI. Consequently, it was considered adequate toaddress the impact of this transition both for the steadystate core analyses as well asfor the stability calculations and to achieve thereby, an integral approach for the validationof the new S3K methodology. Therefore, a comparative assessment of C4 versusC5M is also presented in this paper with particular emphasis on the void coefficientsand their impact on the downstream stability analysis results.5:05 PMEvaluation of the Coolant Reactivity Coefficient Influence onthe Dynamic Response Of A Small LFR SystemS. Lorenzi, S. Bortot, A. Cammi, R. PonciroliDepartment of Energy, Nuclear Engineering Division – CeSNEF, Politecnico di Milano, Milano, ItalyAn assessment of the coolant reactivity feedback influence on a small Lead-cooledFast Reactor (LFR) dynamics has been made aimed at providing both qualitative andquantitative insights into the system transient behavior depending on the sign of theabove mentioned coefficient. The need of such an investigation has been recognizedsince fast reactors cooled by heavy liquid metals show to be characterized by a strongcoupling between primary and secondary systems. In particular, the coolant densityand radial expansion coefficients have been attested to play a major role in determiningthe core response to any perturbed condition on the Steam Generator (SG)side. The European Lead-cooled SYstem (ELSY)-based demonstrator (DEMO) hasbeen assumed as the reference LFR case study. As a first step, a zero-dimensionaldynamics model has been developed and implemented in MATLAB/SIMULINK® environment;then typical transient scenarios have been simulated by incorporating theactual negative lead density reactivity coefficient and its opposite. In all the examinedcases results have shown that the reactor behaves in a completely different way whenconsidering a positive coolant feedback instead of the reference one, the system freedynamics resulting moreover considerably slower due to the core and SG mutuallyconflicting reactions. The outcomes of the present analysis may represent a usefulfeedback for both the core and the control system designers.4:15 PMOn the Effect of Different Placing ZrH Moderator Material onThe Performance of a SFR CoreBruno Merk (1) and Frank-Peter Weiß (2)1) Helmholtz-Zentrum Dresden-Rossendorf, Institut für Sicherheitsforschung, Dresden, Germany, 2) Gesellschaftfür Anlagen- und Reaktorsicherheit (GRS) mbH Forschungszentrum, Garching, GermanyThis study describes the development of a sodium fast reactor fuel assembly designwith reduced void reactivity coefficient, achieved through the use of the ZrH moderatingmaterial. In the study the sodium void effect, as well as the major feedback coefficientsare analyzed. Besides the feedback coefficients, the influence on the operationalparameters like neutron flux distribution, power distribution, and burnup distributionis investigated for the different possibilities of arranging the moderating material in thefuel assembly. Additionally, the fuel cycle parameters – breeding and minor actinideproduction – are analyzed. For a first evaluation of the behavior during transients theinfluence of temperature changes in the ZrH is studied.4:40 PMSpent Fuel Pool Analysis Using Trace CodeF. Sanchez-Saez, S. Carlos, J. F. Villanueva, S. MartorellDepartment of Chemical and Nuclear Engineering, Universitat Politènica de València, València SpainThe storage requirements of Spent Fuel Pools have been analyzed with the purpose toincrease their rack capacities. In the past, the thermal limits have been mainly evaluatedwith conservative codes developed for this purpose, although some works canbe found in which a best estimate code is used. The use of best estimate codes isinteresting as they provide more realistic calculations and they have the capabilityof analyzing a wide range of transients that could affect the Spent Fuel Pool. Twoof the most representative thermal-hydraulic codes are RELAP-5 and TRAC. Nowadays,TRACE code is being developed to make use of the more favorable characteristicsof RELAP-5 and TRAC codes. Among the components coded in TRACE thatcan be used to construct the model, it is interesting to use the VESSEL component,which has the capacity of reproducing three dimensional phenomena. In this work, athermal-hydraulic model of the Maine Yankee spent fuel pool using the TRACE codeis developed. Such model has been used to perform a licensing calculation and theresults obtained have been compared with experimental measurements made at thepool, showing a good agreement between the calculations predicted by TRACE andthe experimental data.5:30 PMThe Solution of the Point Kinetics Equations Via ConvergedAccelerated Taylor Series (CATS)B. Ganapol and P. Picca (1), A. Previti and D. Mostacci (2)1) Department of Aerospace and Mechanical Engineering, University of Arizona. 2) Laboratorio di Montecuccolino,Alma Mater Studiorum - Università di BolognaThis paper deals with finding accurate solutions of the point kinetics equations includingnonlinear feedback, in a fast, efficient and straightforward way. A truncated Taylorseries is coupled to continuous analytical continuation to provide the recurrence relationsto solve the ordinary differential equations of point kinetics. Non-linear (Wynnepsilon)and linear (Romberg) convergence accelerations are employed to providehighly accurate results for the evaluation of Taylor series expansions and extrapolatedvalues of neutron and precursor densities at desired edits. The proposed ConvergedAccelerated Taylor Series, or CATS, algorithm automatically performs successivemesh refinements until the desired accuracy is obtained, making use of the intermediateresults for converged initial values at each interval. Numerical performance isevaluated using case studies available from the literature. Nearly perfect agreementis found with the literature results generally considered most accurate. Benchmarkquality results are reported for several cases of interest including step, ramp, zigzagand sinusoidal prescribed insertions and insertions with adiabatic Doppler feedback. Alarger than usual (9) number of digits is included to encourage honest benchmarking.The benchmark is then applied to the enhanced piecewise constant algorithm (EPCA)currently being developed by the second author.93

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationThursday April 19, 2012 - 3:50 PM - 301 E8H - Advanced Modeling & Simulation in Reactor PhysicsSession Chair: G. Sjoden (GT)3:50 PMA Simplified DEM-CFD Approach for Pebble Bed Reactor SimulationsYanheng Li and Wei JiDepartment of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy,NYIn pebble bed reactors (PBR’s), the pebble flow and the coolant flow are coupled witheach other through coolant-pebble interactions. Approaches with different fidelitieshave been proposed to simulate similar phenomena. Coupled Discrete Element Method-ComputationalFluid Dynamics (DEM-CFD) approaches are widely studied and appliedin these problems due to its good balance between efficiency and accuracy. Inthis work, based on the symmetry of the PBR geometry, a simplified 3D-DEM/2D-CFDapproach is proposed to speed up the DEM-CFD simulation without significant loss ofaccuracy. Pebble flow is simulated by a full 3-D DEM, while the coolant flow field is calculatedwith a 2-D CFD simulation by averaging variables along the annular directionin the cylindrical geometry. Results show that this simplification can greatly enhancethe efficiency for cylindrical core, which enables further inclusion of other physics suchas thermal and neutronic effect in the multi-physics simulations for PBR’s.4:15 PMDevelopment and Preliminary Verification of the 3D CoreNeutronic Code: COCOHaoliang Lu, Kun Mo, Wenhuai Li, Ning Bai, Jinggang LiReactor Design and Fuel Management Research Center, China Nuclear Power Technology ResearchInstitute, Shenzhen, ChinaAs the recent blooming economic growth and following environmental concerns, Chinais proactively pushing forward nuclear power development and encouraging the tappingof clean energy. Under this situation, CGNPC, as one of the largest energy enterprisesin China, is planning to develop its own nuclear related technology in orderto support more and more nuclear plants either under construction or being operation.This paper introduces the recent progress in software development for CGNPC. Thefocus is placed on the physical models and preliminary verification results during therecent development of the 3D COre Neutronic COde: COCO. In the COCO code, thenon-linear Green’s function method is employed to calculate the neutron flux. In orderto use the discontinuity factor, the Neumann (second kind) boundary condition is utilizedin the Green’s function nodal method. Additionally, the COCO code also includesthe necessary physical models, e.g. single-channel thermal-hydraulic module, burnupmodule, pin power reconstruction module and cross-section interpolation module. Thepreliminary verification result shows that the COCO code is sufficient for reactor coredesign and analysis for pressurized water reactor (PWR).5:05 PMNew Approach to Creation of Geometrical Module for NuclearReactor Neutron Transport Computer Simulation AnalysisTamara Poveschenko and Oksana PoveschenkoRRC “Kurchatov Institute”, Moscow, RussiaThis paper presents the new approach to creation of geometrical module for nuclearreactor neutron transport computer simulation analysis so called the differential crossmethod. It is elaborated for detecting boards between physical zones. It is proposedto use GMSH open source mesh editor extended by some features: a special optionand a special kind of mesh (cubic background mesh).This method is aimed into MonteCarlo Method as well as for deterministic neutron transport methods. Special attentionis attended for reactor core composed of a set of material zones with complicate geometricalboundaries. The idea of this approach is described. In general case methodworks for 3-D space. Algorithm of creation of the geometrical module is given. 2-Dneutron transport benchmark-test for RBMK reactor cluster cell is described. It demonstratesthe ability of this approach to provide flexible definition of geometrical meshingwith preservation of curved surface or any level of heterogeneity.5:30 PMMCNP Modeling of the Swiss LWRs for the Calculation of theIn- and Ex-Vessel Neutron Flux DistributionsM. Pantelias, B. Volmert, S. Caruso (1), P. Zvoncek (2), B. Bitterli (3), E.Neukaeter, W. Nissen (4), G. Ledergerber (5), R. Vielma (6)1) National Cooperative for the Disposal of Radioactive Waste (Nagra), Wettingen, Switzerland. 2)Laboratory for Nuclear Energy Systems, ETH Zurich, Zurich, Switzerland. 3) Kernkraftwerk Goesgen-Daeniken AG, Daeniken, Switzerland. 4) BKW FMB Energie AG-Kernkraftwerk Muehleberg, Muehleberg,Switzerland. 5) Kernkraftwerk Leibstadt AG, Leibstadt, Switzerland, 6) Axpo AG-Kernkraftwerk Beznau,Doettingen, SwitzerlandMCNP models of all Swiss Nuclear Power Plants have been developed by the NationalCooperative for the Disposal of Radioactive Waste (Nagra), in collaboration with theutilities and ETH Zurich, for the 2011 decommissioning cost study. The estimation ofthe residual radionuclide inventories and corresponding activity levels of irradiatedstructures and components following the NPP shut-down is of crucial importance forthe planning of the dismantling process, the waste packaging concept and, consequently,for the estimation of the decommissioning costs. Based on NPP specific data,the neutron transport simulations lead to the best yet knowledge of the neutron spectranecessary for the ensuing activation calculations. In this paper, the modeling concepttowards the MCNP-NPPs is outlined and the resulting flux distribution maps are presented.4:40 PMThe Challenges on Uncertainty Analysis for Pebble BedHTGRChen Hao, Fu Li, Han ZhangInstitute of Nuclear and New Energy Technology, Tsinghua University, Beijing, ChinaThe uncertainty analysis is very popular and important, and many works have beendone for Light Water Reactor (LWR), although the experience for the uncertainty analysisin High Temperature Gas cooled Reactor (HTGR) modeling is still in the primarystage. IAEA will launch a Coordination Research Project (CRP) on this topic soon.This paper addresses some challenges for the uncertainty analysis in HTGR modeling,based on the experience of OECD LWR Uncertainty Analysis in Modeling (UAM)activities, and taking into account the peculiarities of pebble bed HTGR designs. Themain challenges for HTGR UAM are: the lack of experience, the totally different codepackages, the coupling of power distribution, temperature distribution and burnup distributionthrough the temperature feedback and pebble flow. The most serious challengeis how to deal with the uncertainty in pebble flow, the uncertainty in pebble bedflow modeling, and their contribution to the uncertainty of maximum fuel temperature,which is the most interested parameter for the modular HTGR.94

PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationFriday April 20, 2012Technical ToursOak Ridge National Laboratory/Y-12 Sites TourThe ORNL site tour will leave from the conference center at 8:30 AM and return at approximately 3 PM.The tour will include the High Flux Isotope Reactor (HFIR), the Radiochemical Engineering Development Center (REDC), the Spallation Neutron Source(SNS), and the ORNL Leadership Computing Facilities.Cost $20 (includes lunch).Oak Ridge National Laboratory (ORNL), nestled in the rolling hills and between the meandering waterways of east Tennessee, is home to some of themost advanced, cutting edge research in the world. With a technical work force of more than 3000 scientists and engineers and an annual budget of $1.4billion, ORNL pioneers the development of new energy sources, technologies, and materials and the advancement of knowledge in the biological, chemical,computational, engineering, environmental, physical, and social sciences.The tours of ORNL being offered in coordination with PHYSOR 2012 will focus on the cutting edge nuclear research facilities.ORNL Leadership Class ComputingORNL is currently home to the world’s 3rd fastest supercomputer, Jaguar. Jaguar utilizes nearly 225,000 computer processors to solve some of the mostdifficult problems in scientific research and is the key workhorse behind the DOE’s Modeling and Simulation of Nuclear Reactors Hub (CASL). We will havea brief tour of the facility and a demonstration of the research being performed on Janguar.Spallation Neutron SourceThe Spallation Neutron Source (SNS) at ORNL is an accelerator based neutron source that provides the most intense pulsed neutron beam in the worldfrom a 1.4 MW proton beam. The SNS main experiment hall will eventually contain 25 one-of-a-kind instruments to provide opportunities for studies inpractically every scientific and technical field.High Flux Isotope Reactor (HFIR)Operating at 85 MW, HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world. The intense neutron flux, constantpower density, and constant-length fuel cycles are used by more than 200 researchers each year for isotope production, material irradiation, and neutronscattering research.Graphite ReactorThe Graphite Reactor, designed for war-time production of plutonium, was built in only 11 months. Its job was to show that plutonium could be extractedfrom irradiated uranium slugs, and its first major challenge was to produce a self-sustaining chain reaction. Workers began loading uranium into the reactorduring the afternoon of Nov. 3, 1943, and progress was swift. Before dawn on Nov. 4, Enrico Fermi was summoned from a nearby guest house. The reactor“went critical” at 5 a.m.; less than two months later, it was producing a third of a ton of irradiated uranium a day. Two months later, Oak Ridge chemistsproduced the world’s first few grams of plutonium.Y-12 National Security ComplexThe Y-12 National Security Complex site tour will consist of a visit to two of the original Manhattan Project buildings where the electromagnetic separationof Uranium 235 for the world’s first atomic bomb used in warfare, Little Boy, took place. In one of the buildings, the actual electromagnets are on displayalong with artifacts of the Cold War. The bird’s eye view from atop Chestnut Ridge allows the ½ mile wide and 2 ½ mile long to be seen including a viewof the Highly Enriched Uranium Materials Facility where all the nation’s highly enriched uranium not in an active weapon is stored. The tour will last approximatelyone and one-half hours. There are museum artifacts on display at the Y-12 History Center and DVD’s as well as brochures are available there.Registrants for this tour will be required to be US citizens, supply their name and last 4-digits of their social security number, and show a photo ID forbadging the day of the tour.Bellefonte Nuclear Power PlantThe Bellefonte tour will leave from the conference hotel at 7 AM and return at approximately 4 PM.The group will tour inside containment and the turbine building.Cost $50 (includes lunch).About Bellefonte Nuclear Power Plant.On Aug. 18, 2011, the Tennessee Valley Authority (TVA) board of directors approved the completion of Bellefonte Nuclear Plant Unit 1 in Hollywood, Ala.When it begins commercial operation between 2018 and 2020, the 1,260 megawatt reactor will be the largest in TVA’s nuclear fleet. Bellefonte’s outputwill be equal to the electricity needs of 750,000 homes. Bellefonte Unit 1 offers a very unique opportunity. Construction was suspended on Bellefonte Unit1 and Unit 2 in 1988, and both remain uncompleted. Neither reactor has been operated; it is possible to enter the reactor containment and see the vastscale of commercial nuclear power plants.95

96PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and EducationAPRIL 15-20, 2012Meeting Rooms

PHYSOR 2012 Program/Proceedings CD-ROMAbout this CD-ROMThe material in this CD-ROM was published using Adobe© technology.Included on the CD-ROM are versions of Acrobat Reader for Microsoft© Windows TM , Apple© Macintosh TM (Mac OS X), and Unix©InstallationTo view files on this CD-ROM you must have Adobe Reader installed on your hard drive. Installation instructions can be found in theREADME.TXT file.Getting StartedWindows users: Software included in this CD-ROM should automatically launch the proceedings. You can always start viewing thecontent by opening the Start.pdf file provided Adobe Reader has been installed on your hard drive.MacOS X and Unix users: To start open the Start.pdf file.Copyright © 2012American Nuclear Society - ANSProgram Book, CD-ROM, WebSite, Online Paper Submission and Review, and Online Registration areservices/products of Techno-Info Comprehensive Solutions.http://techno-info.com

PHYSOR 2012 ProgramSUNDAYWORKSHOPS - Also see listing at bottom.Ballroom A Meeting Room 301 A Meeting Room 301 B Meeting Room 301 C Meeting Room 301 D Meeting Room 301 E Henley8:00 Workshop #8 - Short Course Workshop #4 -Scientificon Uncertainty Quantification & Computing for Nuclear EngineeringSensitivity Analysis MethodsWorkshop #3 -HybridRadiation Transport Methods forWhole-Reactor-Core Analysisand Nuclear SafeguardsApplicationsWorkshop #2 - Reactor PhysicsAnalysis Using SCALE/TRITONWorkshop #7 - AdvancedReactor ConceptsWorkshop #9 - ConvergenceAcceleration and ReactorPhysics10:00Coffee Break10:15 Workshop #8 - Continued Workshop #4 - Continued Workshop #3 - Continued Workshop #2 -Continued Workshop #7 - Continued Workshop #9 - Continued12:0013:30 Workshop #1 - SCALE Sensitivityand Uncertainty AnalysisWorkshop #5 - AdvancedMonte Carlo for Reactor PhysicsCore AnalysisWorkshop #6 - PARCSWorkshop: Nuclear ReactorSimulationWorkshop #7 - AdvancedReactor ConceptsWorkshop #10 - Applicationof Attila for Reactor Physics15:30Coffee Break15:45 Workshop #1 - Continued Workshop #5 - Continued Workshop #6 - Continued Workshop #7 - Continued Workshop #10 - Continued17:3018:30MONDAY8:30AdjournWelcome Reception at the Knoxville Convention CenterBallroom E-GOpening Plenary SessionLunch Break (on your own)Henley ConcourseConcourseConference Registration 12-6 PM,Exhibitor Booths, Coffee BreakArea10:1510:45Coffee BreakOpening Plenary Session, continued12:00Lunch Buffet (Ballroom A-D) (12:00 - 13:30)Meeting Room 301 C Meeting Meeting Room 301 A Meeting Room 301 B Room 301 D Meeting Room 301 E13:30 1A Core Analysis Methods 2A Deterministic Transport Theory 4A Reactor Concepts & Designs SS2A Special Session 2 on 8A Advanced Modeling &Radiation Transport Methods forWhole Reactor Core AnalysisSimulation in Reactor PhysicsConference Registration and Information,Exhibitor Booths, Coffee Break AreaRegistration 7:00 AM - 4:00 PM15:35Coffee Break15:50 1B Core Analysis Methods 2B Deterministic Transport Theory 3A Monte Carlo Methods &DevelopmentsSS2B Special Session 2 onRadiation Transport Methods forWhole Reactor Core Analysis4B Reactor Concepts & Designs17:55Adjourn19:00Evening at the Women's Basketball Hall of Fame (Shuttles to event start at 18:30)Joan Cronan, University of Tennesse Director of Women’s AthleticsTUESDAY Meeting Room 301 A Meeting Room 301 B Meeting Room 301 C Meeting Room 301 D Meeting Room 301 EHenley Concourse8:00 13A Fuel, Material, Mechanical 2C Deterministic Transport Theory 4C Reactor Concepts & Designs SS2C Special Session 2 onAnalysis & BehaviorRadiation Transport Methods forWhole Reactor Core Analysis10:05Coffee Break10:20 1C Core Analysis Methods 2D Deterministic Transport Theory 3B Monte Carlo Methods & 4D Reactor Concepts & DesignsDevelopments12:00Conference Sit-Down Lunch - Ballroom A-D (12:00 - 13:30)Guest Speaker: Jerry Hopwood, Vice President, Marketing and Product Development Candu Energy Inc.8B Advanced Modeling &Simulation in Reactor Physics8C Advanced Modeling &Simulation in Reactor PhysicsConference Registration and Information,Exhibitor Booths, Coffee Break Area13:30 12A Sensitivity & UncertaintyAnalysisSS1A Special Session 1 in honor ofNils Göran Sjöstrand, the 2011winner of the Eugene P. WignerReactor Physicist Award3C Monte Carlo Methods &Developments16A Radiation Applications &Nuclear Safeguards9A Research Reactors &Spallation SourcesRegistration 7:00 AM - 4:00 PM16:25Poster Session - Ballroom E-G (16:25 - 18:00+)18:00Three River Rambler - vintage train adventureShuttles to event start at 17:30. Train departs station at 18:00WEDNESDAY Meeting Room 301 A Meeting Room 301 B Meeting Room 301 C Meeting Room 301 D Meeting Room 301 EHenley Concourse8:00 12B Sensitivity & UncertaintyAnalysis5A Panel on Education: Setting theStage for the Work Force of theFuture – The DOE Next GenerationSafeguards Initiative’s UniversityCurriculum Development9B Research Reactors &Spallation Sources7A Fuel Cycle & ActinideManagement11A Nuclear Data10:05Coffee Break10:20 1D Core Analysis Methods 5B Education in Reactor Physics 3D Monte Carlo Methods &Developments7B Panel on MOX Fuel“Weapons Grade MOX Fuel – A Way15A Experimental Facilities &Experimentsto a Safer Future”12:00Lunch Buffet (Ballroom A-D) (12:00 - 13:30)13:30 1E Core Analysis Methods 4E Reactor Concepts & Designs 3E Monte Carlo Methods & 7C Fuel Cycle & Actinide11B Nuclear DataDevelopmentsManagement15:35Coffee Break15:50 1F Core Analysis Methods 2E Deterministic Transport Theory 6A Reactor Operation & Safety 14A Reactor Transient & SafetyAnalysis8D Advanced Modeling &Simulation in Reactor Physics17:55Adjourn18:30Banquet - Knoxville Convention Center BallroomSpeakers: H. Lee Dodds and Paul Turinsky, Honorary Chairs301 C Meeting Room 301 D THURSDAY Meeting Room 301 A Meeting Room 301 B Meeting Room Meeting Room 301 E8:00 6B Reactor Operation & Safety 15B Experimental Facilities &Experiments14B Reactor Transient & Safety 7D Fuel Cycle & ActinideAnalysisManagement8E Advanced Modeling &Simulation in Reactor Physics10:05Coffee Break10:20 1G Core Analysis Methods 12C Sensitivity & UncertaintyAnalysis3F Monte Carlo Methods &Developments11C Nuclear Data8F Advanced Modeling &Simulation in Reactor Physics12:00 Conference Sit-Down Lunch (Ballroom A-D)Guest Speaker: Mike Houts, NASATo be followed byPHYSOR 2014 – Informational Session; representative of the ANS Reactor Physics Division13:30 1H Core Analysis Methods 12D Sensitivity & UncertaintyAnalysis14C Reactor Transient & SafetyAnalysis15C Experimental Facilities &Experiments8G Advanced Modeling &Simulation in Reactor Physics15:35Coffee Break15:50 10A Nuclear Criticality Safety 12E Sensitivity & UncertaintyAnalysis11D Nuclear Data14D Reactor Transient & SafetyAnalysis8H Advanced Modeling &Simulation in Reactor Physics17:55Adjourn18:00Sunsphere Spectacular - Henley Concourse and walk to the adjacent sunsphere.Meet onConference Registration and Information,Exhibitor Booths, Coffee Break AreaRegistration 7:00 AM - 2:00 PMHenley ConcourseConference Registration and Information,Exhibitor Booths, Coffee Break AreaRegistration 7:00 AM - 2:00 PMFRIDAY7:008:0016:00— Tour to TVA's Bellefonte Nuclear Power Plant— Oak Ridge National Laboratory tourReturn to pick-up spotT O U R S— Y12 National Security Complex tour