RRFM 2009 Transactions - European Nuclear Society

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volume in the material, which affects the fluidity of the fuel and as a result the fission gas diffusivity and swelling behavior. For the U3Si2, which behaves well during irradiation, the increase in free volume during amorphization is negligibly small, and for U3Si, which behaves poorly during irradiation, the increase in free volume during amorphization is relatively large. This means that just because a material goes amorphous it is not a given that there will be a large swelling increase unless there is a significant increase in free volume. The morphology of fission gas bubbles in actual irradiated uranium-silicide fuels have been discussed in References 9 and 10. In [9], it is reported that irradiated USi and U3Si2 fuels develop uniform, round, very small fission gas bubble, while U3Si displays bubbles that have no uniformity and show signs of migration and interlinkage. In [10], it is reported that the size of the fission gas bubbles in irradiated uranium-silicide fuels is related to the composition of the fuel particles. Many very small (100-300 nm) fission gas bubbles are observed in irradiated USi, while irradiated U3Si2 has fewer bubbles with areas that contain larger fission gas bubbles (up to 1 µm). The bubbles observed in these two fuels were perfectly round and showed no indication of bubble coalescence. However, like was the case in [9], the U3Si fuel exhibits unstable fission gas behavior in that the fission gas bubbles vary in shape and exhibit signs of interlinkage. For the case of the Si-containing interaction layers in fuel plate R2R010, the Si may be having a similar effect as it did for the uranium-silicide fuels. For areas of the interaction layers where the Si content is relatively high, the amount of free volume increase is small when the material goes amorphous during irradiation, and for cases where the Si content is relatively low the free volume increase is larger and the fission gas bubbles can grow and become more mobile. In the R2R010 punchings that were analyzed using SEM [5], it was actually observed that the fission gas bubbles significantly increased in size in some interaction layer regions where the Si concentration was relatively low. This would suggest that the properties of the interaction layer had changed because of the lower amount of Si (perhaps the free volume), which resulted in a lower viscosity of the amorphous material and a resulting growth of the bubbles and an increase in the mobility of the fission gases. However, since the fission gas bubbles were still contained in the interaction layer and had not migrated to the interaction layer/matrix interface, there still seemed to be enough Si present in the layer to keep the viscosity high enough to keep the bubbles in the layer. 5. Conclusions Based on the TEM characterization of a U-7Mo/Al-2Si dispersion fuel plate irradiated to moderate burnup in the RERTR-6 experiment, the following conclusions can be drawn: 1. U-7Mo fuel particles irradiated to medium burnup retain their crystallinity and contain small fission gas bubbles distributed on a super-lattice. 2. The Si-rich interaction layers that originally develop around U-7Mo particles during fuel plate fabrication and are present when a fuel plate is inserted into the reactor become amorphous during irradiation. These layers exhibit stable irradiation behavior probably due to the fact that they appear to retain fission gases as small,
References [1]. D. Wachs, et al., Proc. of GLOBAL 2007, Boise, ID, September 9-13, 2007. [2]. C. R. Clark et al., <strong>RRFM</strong> 2004, Munich, Germany, March, 2004. [3]. S. Van den Berghe et al., J. Nucl. Mater. 375 (2008) 340-346. [4]. G. L. Hofman et al., RERTR 2004, Vienna, Austria. [5]. D. D. Keiser, Jr. et al., this proceedings. [6]. J. Gan et al., Hot Laboratories and Remote Handling Conference (HOTLAB 2008), Kendal, England, Sep. 22-23, 2008. [7]. D. D. Keiser, Jr. et al., RERTR 2008, Washington D. C. [8]. G. L. Hofman and Y. S. Kim, Nucl. Engr. and Tech. 37 (2005) 299-308. [9]. M. R. Finlay et al., J. Nucl. Mater. 325 (2004) 118-128. [10]. A. Leenaers et al. J. Nucl. Mater. 375 (2008) 243-251. 439 of 455
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© 2009 European Nuclear Society Ru
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RESEARCH REACTOR COALITIONS - SECON
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- Eastern European Research Reactor
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ut a comprehensive alignment of tec
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EERRI will be launched. As noted ab
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Nuclear material in the form of hig
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The GTRI domestic radiological mate
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2. State of the art For fuel plate
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It is known since the early days of
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End of 2008 first DU-8wt.%Mo (deple
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References [1] D. AB. Robinson et a
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• CNEA have been working in the f
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O 75 Bignan.doc - DI - 1 / 10 20/02
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THE EAST EUROPEAN RESEARCH REACTOR
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namely Jozef Stefan Institute TRIGA
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Information and organizational issu
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Action Item Table 3. Description (t
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Table 6. Materials/fuel test experi
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In a recent development, it should
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1. Introduction: a renewed context
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interaction, close retention of fis
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3.3 Reference concept and alternati
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changes can be caused by void swell
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For its manufacturing, an additiona
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minor actinides (MA) considered as
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In France, the decision to build a
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eactors (HFR, OSIRIS and then JHR)
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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 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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Page 396 and 397: etween CRPq specific process and IP
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Page 402 and 403: MCNP5 is capable of calculating ter
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Page 406 and 407: LEU FUEL DISPOSITION AT WWR-M REACT
Page 408 and 409: Tested fuel assembles are load in t
Page 410 and 411: The PNPI experience of tests with W
Page 412 and 413: CURRENT UTILIZATION AND LONG TERM S
Page 414 and 415: Uusimaa hospital district decided t
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Page 418 and 419: STUDYING OF INFLUENCE OF A NEUTRON
Page 420 and 421: Definition of crystal structure and
Page 422 and 423: Microhardness measurement shows, th
Page 424 and 425: THE SPENT FUEL STORAGE AT THE DALAT
Page 426 and 427: 2. Interim storage of spent fuel On
Page 428 and 429: A rotating bridge crane of 3.6-ton
Page 430 and 431: In order to investigate the perform
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Page 434 and 435: (a) (b) (c) (d) (e) (f) Fig. 5. TEM
Page 436 and 437: Spot Al Si Mo U Zr G 90.6 7.8 0.6 0
Page 440 and 441: THE STEREOMETRIC ANALYSIS OF GAS PO
Page 442 and 443: of the crack of pillowing). It show
Page 444 and 445: One gets the impression that during
Page 446 and 447: switch from a fast to a thermal spe
Page 448 and 449: He production to fission ratio vs t
Page 450 and 451: INSPECTION EXPERIENCE WITH IEA-R1 S
Page 452 and 453: camera system, received from IAEA i
Page 454: core in the beginning 2007 (IEA-174
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