RRFM 2009 Transactions - European Nuclear Society

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Table 6. Materials/fuel test experiments (loop/rigs) operated in RRs of EERRI Coalition Reactor LVR 15 (CZECH REPUBLIC) (Country) Facility BWR 1 BWR 2 RVS 3 RVS 4 CHOUCA FLAT IRRADIATION RIG Type in-pile loop in-pile loop in-pile loop in-pile loop irradiation rig irradiation rig Purpose material behaviour and radioactivity transport material behaviour and radioactivity transport neutron irradiation of constructional materials under BWR conditions under PWR/VVER conditions used for reactor vessel construction Parameters medium water water water water He / N / Ar He / N / Ar pressure 11 MPa 12 MPa 16.5 MPa 15.7 MPa 100 kPa 100 kPa temperature 310 °C 310°C 345°C volume 62 l 510 l 210 l flow 3000 kg/hr 3000 kg/hr 10000 kg/hr heat flux / heating capacity 45 kW 100 kW 311-322°C 300 °C 300 °C 10 l 30 l 30 l 2000 kg/hr 60 W/cm 2 , heated length 560 mm 6 x 2 kW 8x800 W / 6x400 W Neutron flux: ~1x10 18 n/m 2 s ~1x10 18 n/m 2 s ~1x10 18 n/m 2 s ~1x10 18 n/m 2 s ~1x10 18 n/m 2 s ~1x10 18 n/m 2 s Specimen Æ 56 x 400 mm 50x120x500 mm, specimen strained space 20x60x260 mm Services • Investigation of materials mechanical • Investigation of structural materials • Tensile specimen, • Charpy-V properties degradation and corrosion behaviour mechanical properties degradation and CT specimen, round specimens under irradiation and BWR water chemistry corrosion behaviour under irradiation and Cts, up to 40 Charpy-V conditions PWR/VVER water chemistry and thermalhydraulic specimens • Investigation of radioactivity transport and conditions behaviour under BWR conditions (eg. hydrogen • Investigation of behaviour (corrosion, water chemistry, zinc injection, etc.) hydriding) of fuel cladding materials under • Testing of high-temperature, high pressure influence of irradiation, thermal flux and water sensors for water chemistry monitoring chemistry conditions MARIA (POLAND) NTD silicon doping 5 and 6 inches 9 44 of 455
Table 6. (Cont.) Reactor BRR (HUNGARY) Triga II Pitesti – SS Core (ROMANIA) (Country) Facility BAGIRA 1 BAGIRA 2 Loop A C1&C2 C5 C6 C9 Type in-pile irradiation rig in-pile irradiation rig loop In pile capsules In pile capsule In pile capsule In pile capsule neutron irradiation of constructional neutron irradiation of constructional Irradiation tests of fuel elements and Irradiation tests of fuel elements Structural materials irradiation tests in CANDU type fuel element tests in fast Cycling tests on fuel elements Purpose materials used for reactor vessel construction materials used for reactor vessel construction structural materials used in PHW reactors Two independent capsules for parametric testing inactive environment transient regimes in TRIGA ACPR reactor Parameters medium He/Nitrogen Demineralised water Helium Demineralised water pressure 300 kPa 300 kPa 13.5 MPa 12 MPa 0.6 MPa 0.4 MPa 10.7 MPa temperature 150-500 °C 70-150 °C 3100 °C 3300 °C on fuel clad 2900 °C 500 °C 3250 °C on fuel clad volume 5 l 5 l 252 l 30 l - convection 7.5 l 3 l flow - - 3-7 m 3 /h 4 m 3 /h - Stagnant water 0 – 4 m 3 /h heat flux / 20.000 MW peak 21.5 kW 80 W - 100 kW 30 kW 10 kW heating capacity power pulse Neutron flux: 4x10 9 n/m 2 s 3x10 9 n/m 2 s ~3.2 x10 18 n/m 2 s ~2x10 18 n/m 2 s ~10 17 n/m 2 s ~2x10 18 n/m 2 s ~2x10 18 n/m 2 s Specimen space 20x30x300 20x20x300 Services • Charpy-V specimens, tensile specimens, CT specimens • Charpy-V specimens, tensile specimens, CT specimens • Overpower type tests on fuel element • Power ramp type tests on fuel element • Corrosion and mechanical behavior studies on structural materials used in CANDU pressure tubes • LOCA type tests • Fuel element dimensional measurement • Fission products pressure – on line • Power ramp • Short-time irradiation for residual deformation of the cladding determination • Structural materials irradiation tests in inactive environment: Zircalloy-4, steel 403- M, Zr-2,5%Nb until 2,3X1022 nvt • Irradiation and tensile test of Chorpy standard minisamples – maximum 30 samples per irradiation campaign • Thermomecanical behavior of CANDU type fuel element in fast power transients • Cycling tests on fuel elements that should confirm the fuel capacity to support a wide range of power cycling that occurs in normal operation of a CANDU reactor during power load following. 10 45 of 455
Page 2 and 3: © 2009 European Nuclear Society Ru
Page 4 and 5: RESEARCH REACTOR COALITIONS - SECON
Page 6 and 7: - Eastern European Research Reactor
Page 8 and 9: ut a comprehensive alignment of tec
Page 10 and 11: EERRI will be launched. As noted ab
Page 12 and 13: Nuclear material in the form of hig
Page 14 and 15: The GTRI domestic radiological mate
Page 16 and 17: 2. State of the art For fuel plate
Page 18 and 19: It is known since the early days of
Page 20 and 21: End of 2008 first DU-8wt.%Mo (deple
Page 22 and 23: References [1] D. AB. Robinson et a
Page 24 and 25: • CNEA have been working in the f
Page 26 and 27: O 75 Bignan.doc - DI - 1 / 10 20/02
Page 36 and 37: THE EAST EUROPEAN RESEARCH REACTOR
Page 38 and 39: namely Jozef Stefan Institute TRIGA
Page 40 and 41: Information and organizational issu
Page 42 and 43: Action Item Table 3. Description (t
Page 46 and 47: In a recent development, it should
Page 48 and 49: 1. Introduction: a renewed context
Page 50 and 51: interaction, close retention of fis
Page 52 and 53: 3.3 Reference concept and alternati
Page 54 and 55: changes can be caused by void swell
Page 56 and 57: For its manufacturing, an additiona
Page 58 and 59: minor actinides (MA) considered as
Page 60 and 61: In France, the decision to build a
Page 62 and 63: eactors (HFR, OSIRIS and then JHR)
Page 64 and 65: DECOMMISSIONING PROGRESS OF THE DOU
Page 70 and 71: SECURITY OF SUPPLY FOR FISSION MEDI
Page 72 and 73: But the worst crisis developed by e
Page 74 and 75: IRE and COVIDEN are following close
Page 76 and 77: Fig 1. Design of the core and refle
Page 78 and 79: Table 3: Set of detailed fuel funct
Page 80 and 81: For the thermal-mechanical analysis
Page 82 and 83: Behaviour under conditions represen
Page 84 and 85: 6. References [1] - MC. Anselmet, G
Page 86 and 87: 2. Advanced nuclear fuel cycles The
Page 88 and 89: 3.3 India In addition to the ration
Page 90 and 91: Session II Fuel Development 90 of 4
Page 92 and 93: 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 5 of 7 programs for instrument
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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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