RRFM 2009 Transactions - European Nuclear Society

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The GTRI domestic radiological material removal program is working in cooperation with Federal, state and local agencies, and private industry to recover and permanently dispose of excess radiological sources in the United States. A cumulative total of over 20,300 domestic sources have been recovered. GTRI is committed to remove at least 2,500 excess domestic radiological sources each year. While this is an impressive milestone, each year over 3,000 new sources are registered as excess creating a backlog of more that 8,000 sources to recover. 2.3 The Protect Program GTRI’s Protect Program involves both international and domestic material protection. Work is conducted to ensure material security building by building. Many of the buildings holding nuclear and radiological materials require a different approach since they are accessible to the public, such as hospitals, and university facilities. A systematic approach is applied to evaluate and implement security measures. The GTRI team works with international and domestic partners to: perform site protection assessments; design the security upgrades; obtain a site sustainability commitment; install the security upgrades; and conduct table top exercises with facility staff to ensure an understanding and proficiency with the protection upgrade. Working with Federal, state and local agencies, GTRI has established a domestic goal of 2,191 high-priority U.S. buildings that require protection. This work has recently begun, and 17 have already been completed International nuclear and radiological material protection has identified 1,759 buildings that require protection. To date, 30% are completed. Acceleration of this effort is particularly important because upgraded security is necessary until a permanent threat reduction solution can be implemented 3. Conclusion Secretary of State Hillary Clinton stated in her confirmation hearing on January 13, 2009, “……The gravest threat ......is the danger that weapons of mass destruction will fall into the hands of terrorists. To ensure our future security, we must curb the biological, chemical, or cyber — while we take the lead in working with others to reduce current nuclear stockpiles and prevent the development and use of dangerous new weaponry….” iii This urgent warning has been echoed by many other WMD experts and expert committees. In order to meet President Obama’s goal to secure nuclear weapons material in four years, the NNSA has developed a plan to accomplish this goal. Every facet of the GTRI mission will contribute to the success of this plan. International partners around the world must be involved and will be welcomed in these important efforts. 4. References i White House website, www.white.house.gov/agenda/homeland_security/ ii Ibid. iii Testimony of Senator Hillary Clinton before the Senate Foreign Relations Committee, January 13, 2009. 14 of 455
FRM II AND AREVA-CERCA COMMON EFFORT ON MONOLITHIC UMO PLATE PRODUCTION C. JAROUSSE, L. HALLE AREVA-CERCA - Les Berauds, B.P. 1114, 26104 Romans Cedex – France W. PETRY, R. JUNGWIRTH, A. RÖHRMOSER, W. SCHMID Forschungsneutronenquelle Heinz Maier-Leibnitz (FRM II) Technische Universität München, D-85747 Garching, Germany ABSTRACT Since 2003 the Technische Universität München (TUM) is strongly engaged on the UMo fuel development program. Dispersive UMo fuel type was firstly investigated through collaborative efforts involving CEA, TUM and AREVA-CERCA. IRIS TUM irradiation and associated PIEs were presented during international conferences. The European consortium also studied the fabrication of UMo monolithic fuel plates during 2005-2007 and tentatively manufactured full size plates for the irradiation program IRIS V. As the program unfolded technical information were obtained and gathered with the international community. UMo foils were produced at laboratory scale and different methods to clad the UMo foils within aluminium were investigated. Based on these first results TUM and AREVA-CERCA have decided to pursue a common effort on the development of monolithic fuel plates with the purpose of minimizing the enrichment in use. Emphasis will be more dedicated to the foil and cladding package preparation and further studies will be performed in order to investigate various processing techniques to join the fuel foil with the cladding. As a complete new approach for the manufacturing of monolithic foils and for foil cladding the DC magnetron sputtering technique is investigated. A full R&D program was defined between TUM and AREVA-CERCA. This paper aims at presenting the program, discusses the selected options and first results will be presented 1. Introduction The monolithic fuel concept presented by INL early in 2002 allows the possible LEU conversion of high performance reactors which can’t be converted using dispersed UMo fuel. The basic principle is to replace a dispersive fuel form with a density of 8 gU/cc by a solid form of UMo where a density close to 17 gU/cc can be reached. The technical challenges for this kind of fuel are multiplex: Instead of having numerous interfaces between the matrix and each grain the interface is now reduced to two large interfaces on top and the bottom of the monolithic foil. Which kind of interlayer is best suited for an excellent adhesion of the cladding and for suppressing the interdiffusion layer? How to produce large foils with a typical thickness of 300 – 500 µm on an industrial scale to affordable costs? How to introduce gradients in thickness for the foils? Can the assembled fuel plate be shaped to geometries others than flat? And last but not least, how do large tests plates behave under in-pile irradiations which simulate the heat load and burn-up of high performance reactors? AREVA-CERCA, a subsidiary of AREVA NP, an AREVA and SIEMENS company 15 of 455
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DECOMMISSIONING PROGRESS OF THE DOU
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SECURITY OF SUPPLY FOR FISSION MEDI
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But the worst crisis developed by e
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IRE and COVIDEN are following close
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Fig 1. Design of the core and refle
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Table 3: Set of detailed fuel funct
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For the thermal-mechanical analysis
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Behaviour under conditions represen
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6. References [1] - MC. Anselmet, G
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2. Advanced nuclear fuel cycles The
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3.3 India In addition to the ration
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Session II Fuel Development 90 of 4
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1. Introduction Since 1957, date of
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A 3.1. Place a boron insert in a hi
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Finally, a new tool was defined and
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the end user (CEA), to define the f
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Previous papers discussed character
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(a) (b) (c) (d) (e) (f) (g) (h) (i)
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The results of a third reported irr
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In order to reproduce the heat tran
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25 OXIDE THICKNESS 20 Kim et al pH=
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Heavy ion irradiation of UMo7/Al fu
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For the lowest flux values tested d
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127 I beam 127 I beam 0 5 10 15 20
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Weight fractions (%) U(α) UO 2 γU
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activities to meet the need. The co
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Heat capacity Information Thermal c
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Decision point Activity Phase 2: Fu
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IRIS PROGRAM: IRIS4 FIRST RESULTS F
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The main characteristics of the IRI
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4. In-pool plate thickness measurem
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At that stage of IRIS4 irradiation
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extrapolated thermal property value
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optical microscope, acoustic emissi
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CD WIRES AS BURNABLE POISON FOR THE
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Figure 1. MCNP geometry model of fu
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considered carefully: (i) maintain
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discussed and validated together, k
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The as-fabrication Si contents in t
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uniform Si diffusion and consequent
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Recently, fuel relocation induced b
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characteristics of ILs obtained in
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- Type 2 (Si content ≥ 5 wt%) : c
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4. Discussion The main contribution
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METHODS OF INCREASING THE URANIUM C
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• strength properties of Zr-1% Nb
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4. References 1. Birzhevoy G.A., Ka
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Dispersion fuel Monolithic fuel ATR
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− because of its Newtonian charac
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Session III Back-end 166 of 455
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Fig.1. The transfer hall (left in c
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The shipment included all the SNF f
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FRESH AND SPENT NUCLEAR FUEL REPATR
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modifications, spent fuel inspectio
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The shipment cleared Romanian Custo
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steel are chosen as effective gamma
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Fig 3. Long lived nuclides radioact
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Session IV Innovative Methods in Re
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Lower Gr-reflector Al-clad Upper Gr
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All described fuel elements were ca
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ANALYSIS OF NEW SAFETY CRITERIA FOR
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3.2 Strain limit The value of 3.65%
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The comparison between clad tempera
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VALIDATION OF PLTEMP/ANL V3.6 CODE
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ΔT p 0.0234 2 ⎪⎧ q[W/cm ] ⎪
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Table II. Hot Channel Factors for B
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for the U-8Mo fuel meat [5]. The po
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correlation of Dittus-Boelter; the
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ROLE OF RELAP/SCDAPSIM IN RESEARCH
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SEVERE REACTIVITY INJECTION ACCIDEN
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expansion and steam formation were
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3 Conclusion Since a few years, IRS
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analysis procedure (ENAA) except vi
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References Briesmeister, J.F., 2000
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Fig. 1 A geometric diagram of NIRR-
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Fuel plate temperatures during oper
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the primary circuit in the central
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Fig. 3: Surface temperature over th
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[4] “Test Irradiations of Full Si
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The core is made of 1488 stainless
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4. Commissioning the facility for t
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ESTABLISHMENT OF A SINGLE-CRYSTAL F
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were provided to MU and INL for thi
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concrete covered with sheets of 1.2
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USE OF FISSION RADIATION IN LIFE SC
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Fig. 3. Scan of an etch track film
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Summary The unique fission neutron
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2. Methods 2.1 Reactor produced rad
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3.2 Viability studies The results o
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DESIGN OF A FLOWING-NAK EXPERIMENTA
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4. Containment rig and sample-holde
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7. Electrical heater The external h
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Page 1 / 7 EVITA: a semi-open loop
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Page 3 / 7 Fig.1: EVITA fuel Genera
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Page 5 / 7 Challenges When operatin
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Page 7 / 7 References: [1] M.C. Ans
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Nuclear Research Institutes, and so
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According to the latest IAEA inform
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18. May Training of operating perso
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2.1 Irradiation Facilities Brief te
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3.1 Detection limits Although it is
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PARTIAL DISMANTLING OF RESEARCH REA
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• Core - liable to full scale rep
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• Necessary individual dosimetry
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SILICON DOPING AT FRM II H. GERSTEN
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The resulting neutron flux density
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4. References [1] Wagner F.M., Knes
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The choice of fuel base and solutio
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3.3 Increasing power beyond current
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A COATING TO PROTECT SPENT ALUMINIU
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The treatment consisted of simple i
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5. Conclusions 1. Immersion of AA 1
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KEEPING AGING RESEARCH REACTORS IN
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inspected in presence of an expert
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After the inspection was finished,
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1 A STRATEGY FOR MANAGING AGEING CO
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3 2. With the aluminium-clad HEU fu
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5 The pool and the support structur
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Fig.2. Reflector element of JRR-4.
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Dimensional change (%) 0.6 0.5 0.4
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SAFETY CONSIDERATIONS FOR CORE MANA
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3 3. Core operation and monitoring
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5 temperature, etc.). Effective sec
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1436 keV, Cs 138 1369 keV, Na 24 2.
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compared with that of the delayed n
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AD-HOC AND PREVENTATIVE MAINTENANCE
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Page 3 of 7 3. SAFARI-1OPERATIONAL
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Page 7 of 7 DATE DESCRIPTION DATE D
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MICROSTRUCTURAL ANALYSIS OF MTR FUE
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In sample S3, similar zones could b
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increased temperature, the solid st
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Figure 8 BEI images covering Sample
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660 °C, but since in the former me
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support. Both Governments have eval
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• Removal of sludge from the SNF
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UPGRADED REACTOR SYSTEMS FOR ENHANC
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Parameter HEU LEU a 6.0x10 -5 (°K)
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From 1992 until 2005 the TRIGA-SSR
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Figure9 - Dependency between reacto
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URANIUM CONTAMINATION IN PRIMARY CI
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1.00E+06 1.00E+05 Volume activity (
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6. Acknowledgement This work was pe
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2. Steady State Thermal Hydraulic A
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in the scram, and the respective re
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CORROSION BEHAVIOR OF ALUMINIUM ALL
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3.2. Pitting corrosion of 1050 and
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4. Discussions This study shown tha
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The Steady State core was fully con
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0.E+00 2.E+06 1.40E+07 2.E+06 1.20E
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SYNTHESIS AND CHARACTERIZATION OF G
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0 2 4 6 8 10 120 1,2 100 1,0 80 70
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Element, Wt %, At %, K‐Ratio, Z,
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U/cc, but the target uranium densit
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Fig. 5 Macroscopic cutting view (x
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SELF-ASSESSMENT OF APPLICATION OF T
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2. The Code of Conduct on the Safet
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For emergency preparedness, conside
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inside of a compaction die by blowi
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Fig. 4. A compacted and sintered lu
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THE MANAGEMENT SYSTEM EVOLUTION OF
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etween CRPq specific process and IP
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aspect audits, mainly in relation t
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1 2 3 4 5 6 7 8 9 A B C D E F G H F
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MCNP5 is capable of calculating ter
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Similar to the case of stainless st
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LEU FUEL DISPOSITION AT WWR-M REACT
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Tested fuel assembles are load in t
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The PNPI experience of tests with W
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CURRENT UTILIZATION AND LONG TERM S
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Uusimaa hospital district decided t
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FiR 1 has an important regional rol
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STUDYING OF INFLUENCE OF A NEUTRON
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Definition of crystal structure and
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Microhardness measurement shows, th
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THE SPENT FUEL STORAGE AT THE DALAT
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2. Interim storage of spent fuel On
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A rotating bridge crane of 3.6-ton
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In order to investigate the perform
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(a) (b) Fig. 2. (a) low and (b) hig
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(a) (b) (c) (d) (e) (f) Fig. 5. TEM
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Spot Al Si Mo U Zr G 90.6 7.8 0.6 0
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volume in the material, which affec
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THE STEREOMETRIC ANALYSIS OF GAS PO
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of the crack of pillowing). It show
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One gets the impression that during
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switch from a fast to a thermal spe
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He production to fission ratio vs t
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INSPECTION EXPERIENCE WITH IEA-R1 S
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camera system, received from IAEA i
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core in the beginning 2007 (IEA-174
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