RRFM 2009 Transactions - European Nuclear Society

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In France, the decision to build a prototype of 4 th generation fast neutron reactors (ASTRID) by 2020 is important for Europe to remain credibly involved in R&D on fast neutron systems after Phenix is shutdown in 2009 (Figure 6). ASTRID Phenix Figure 6: Phenix and ASTRID in the Marcoule site (artistic view) Other European countries, and possibly other international partners, may decide to develop experimental or prototype reactors of other Generation IV systems of specific interest, with invited external participation. The prospect of such prototypes does not appear excessively ambitious in comparison with the number of experimental and prototype reactors operating in Europe in the 1980s. 5.3 Irradiation programmes SFR fuel development Three main domains are currently demanding for irradiation experiments: the start-up fuel for ASTRID (large diameter oxide fuel pins with a classical austenitic clad), the final driver fuel for ASTRID (large diameter oxide fuel pins with an ODS clad) and the development of transmutation fuel (MAs homogeneous recycling in which a little amount of actinides is diluted in the fuel, and MAs heterogeneous recycling for which the preferred option is currently to concentrate up to 20% of MAs in a UO 2 matrix in radial blanket S/As). At the current stage of development, SFR needs on carbide fuel are considered to be covered by the GFR program. These different options are obviously not at the same level of maturity, therefore requiring to developing different types of irradiation experiments. In any case, after Phénix reactor shutdown in 2009, the need to realize prototypic irradiations on the most mature options will be a difficulty in the 2012-2020 period and France considers that discussions must be internationally opened to equip for example the Monju reactor with irradiation capsules. France is also considering that irradiation capsules are to be developed for the ASTRID prototype. For less mature options, existing MTRs and, after 2014, the Jules Horowitz Reactor (JHR) will have a key role to play to provide analytical and instrumented experiments for acquiring basic knowledge, screening options and providing features on the fuel integral behaviour prior to the realization of prototypic experiments. These general features of the SFR irradiation program are shown on Figure 7. 60 of 455 14/17
Standard oxide fuel with austenitic clad 2008 2012 2020 PHENIX COPIX JOYO, MONJU? Small austenitic cladded bundle Co-precipitated powder ASTRID PHENIX JHR Whole core CALMOS, FACTO PREFACTO Analytical Power to melt experiments Experiments Standard oxide fuel with ODS clad PHENIX MATRIX 1&2 ODS samples JOYO, BN600? ODS samples, pins High doses JOYO, MONJU? Small ODS cladded bundle ASTRID One subassembly with ODS cladding Homogeneous recycling MA bearing fuels Heterogeneous recycling HFR MARIOS JOYO MONJU Several small pins 1 pin GACID Programme JHR, JOYO, MONJU? OSIRIS Analytic and integral DIAMINO experiments ASTRID, MONJU Several pins to one subassembly ASTRID One or several pins Figure 7: SFR irradiation programme ASTRID start-up fuel and driver fuel The objective is to start the ASTRID core with subassemblies made of large diameter oxide fuel pins cladded with a classical austenitic steel and to introduce, very soon after ASTRID start-up, one sub-assembly with ODS cladded pins. The irradiation program takes benefit from the considerable experimental feedback from Rapsodie, Phenix (in which still relevant experiences have been made but are not already examined) and SPX. Major mid-term objective would be to irradiate in prototypic conditions small bundles of large diameter pins: this raises the question of fast neutron reactor availability with appropriate irradiation devices. As far as ODS are concerned, the objective is to acquire quickly high fast neutron fluences on samples of the various candidate materials and then to irradiate the selected ODS cladded pins in SFR representative conditions. The Phenix reactor last neutrons have been used in particular for material irradiation experiment (MATRIX 1 and 2) to start acquiring data on ODS and to check the behaviour of oxide fuel manufactures with (U,Pu) co-precipitated powder. A power-to-melt experiment on pre-irradiated oxide fuels will also be performed in order to enrich the corresponding database necessary for fuel codes qualification for an accurate evaluation in the design calculations of the margins to melting. Transmutation fuels For the homogeneous recycling which is a mature option after the SUPERFACT experiment performed in Phénix in the 80's that has shown a fuel behaviour similar to standard one, the program is performed within the GACID R&D project of the Gen IV collaboration. It has the objective to irradiate MAs bearing fuel in Monju, first at the pin level, then at a more significant scale (several pin or whole subassembly in Monju or GACID). For the heterogeneous recycling and in particular the case of blankets loaded with MAs, the high MAs content raises the question of the management of the large quantity of He produced associated to very particular irradiation parameters evolution. Therefore, a specific transmutation fuel microstructure is to be developed which requires several steps in MTRs 61 of 455 15/17
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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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Page 14 and 15: The GTRI domestic radiological mate
Page 16 and 17: 2. State of the art For fuel plate
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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
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Page 36 and 37: THE EAST EUROPEAN RESEARCH REACTOR
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Page 48 and 49: 1. Introduction: a renewed context
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Page 56 and 57: For its manufacturing, an additiona
Page 58 and 59: minor actinides (MA) considered as
Page 62 and 63: eactors (HFR, OSIRIS and then JHR)
Page 64 and 65: DECOMMISSIONING PROGRESS OF THE DOU
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Page 72 and 73: But the worst crisis developed by e
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Page 76 and 77: Fig 1. Design of the core and refle
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Page 80 and 81: For the thermal-mechanical analysis
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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
Page 94 and 95: A 3.1. Place a boron insert in a hi
Page 96 and 97: Finally, a new tool was defined and
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Page 100 and 101: Previous papers discussed character
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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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