RRFM 2009 Transactions - European Nuclear Society

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switch from a fast to a thermal spectrum in the parameters above can at least be partially counteracted. Methodology For this preliminary parametric study it was decided to use the depletion code DARWIN 2.2 [3] to handle both the JHR [2] case and the SFR [1] reference sample with the same program and data files. As a reference sample serves Am 0.1 U 0.9 O 2 because it’s a likely candidate for the actual experiment and has plenty of margin to change the fission rate by 235 U enrichment. Additionally similar compositions have already been investigated as transmutation/fertile blanket in the SFR core [3]. Starting from the reference sample U enrichment, neutron flux level and Am content, but not isotopic composition, were modified in the JHR spectrum to approach the behaviour of the same sample in the SFR. Table 1 shows reference case for both reactor spectra. SFR core JHR reflector max flux [n/s/cm²] 2.4667x10 15 5.1267x10 14 spectrum ERANOS 33 gr. TRIPOLI 172 gr. reference sample Am 0.1 U 0.9 O 2 density [g/cm³] 10.3 U mass composition Am mass composition 0.25% 235 U, 99.75% 238 U 95% 241 Am, 5% 243 Am Tab 1: Reference cases for JHR and SFR The necessary neutron spectra were provided by existing detailed TRIPOLI4.4 and ERANOS models for the JHR, SFR and PHENIX and are shown below in figure 1. Neutron Spectrum 1.E+15 1.E+14 1.E+13 1.E+12 JHR reflector SFR in core 1.E+11 1.E+10 1.E+09 1.E- 09 1.E- 08 1.E- 07 1.E- 06 1.E- 05 1.E- 04 1.E- 03 1.E- 02 1.E- 01 1.E+00 1.E+01 1.E+02 Energy [MeV] Fig 1: Neutron spectrum used in the calculations 446 of 455
Results The burn up calculations were done in 100 steps with 45 days duration but the results get continuously more uncertain after the first third of the total burn up period especially for the thermal spectrum due to limited self shielding capabilities of DARWIN. This was considered acceptable for the preliminary study in preparation for more detailed and time consuming calculations with HORUS code package. Nevertheless the results demonstrate that matching He to fission ratio and burn up vs. time is possible between fast and thermal spectra. The following three graphs show the behaviour of selected sample and flux combinations. He production 1.40000E-03 1.20000E-03 He mass/ intitial HM mas 1.00000E-03 8.00000E-04 6.00000E-04 4.00000E-04 2.00000E-04 0.00000E+00 0 180 360 540 720 900 1080 1260 1440 1620 1800 1980 2160 time [days] 2340 2520 2700 2880 3060 3240 3420 3600 3780 3960 4140 4320 4500 SFR reference JHR reference JHR 20% enriched, 3% flux JHR 2.5% Am, 33% flux JHR 10% enrichted, 33% flux Fig 2: He content in the sample. Please note flux percentage given for JHR samples refers to the JHR reference flux. Because the reaction chain 241 Am (n,γ) 242 Am (d,β) 242 Cm (d,α) 238 Pu is responsible for most of the He production, the He content is only dependent on the neutron flux level and Am content of a sample for a given neutron spectrum. This means U enrichment can be used to shift the He to fission ratio curve given in figure 3 up or down without changing its shape. While allowing for a very good match between for example the SFR reference and the JHR 10% enriched sample at 33% max. flux, the energy and He deposition per time differ considerable and the difference is burn up dependent. To speed up an experiment from 10 years irradiation time to 5 years for the same burn up is not inconvenient, but the higher power also effects fuel behaviour. On other hand the He contends and burn up curves for the JHR 20% enriched sample with 10% max. flux are individually far closer to their counter parts of the SFR but the He to fission ratio in figure 3 is actually worse than most other examples. 447 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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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
Page 396 and 397: etween CRPq specific process and IP
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Page 402 and 403: MCNP5 is capable of calculating ter
Page 404 and 405: Similar to the case of stainless st
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
Page 416 and 417: FiR 1 has an important regional rol
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
Page 432 and 433: (a) (b) Fig. 2. (a) low and (b) hig
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 438 and 439: volume in the material, which affec
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 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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