RRFM 2009 Transactions - European Nuclear Society

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Dimensional change (%) 0.6 0.5 0.4 0.3 0.2 Dimensional change 0.1 Fast neutron fluence 0 0 200 400 600 Distance from top of graphite reflector (mm) 1.0 0.8 0.6 0.4 0.2 0.0 Fast neutron fluence (×10 24 n/m 2 ) Fig.4 (a). Comparison of the cross-section average between the dimensional change and fast neutron fluence as a function of the distance from top of the graphite reflector (In the case that the irradiation growth was within the gap size). 2.5 6.0 Dimensional change (%) 2 1.5 1 0.5 0 Dimensional change Fast neutron fluence 0 200 400 600 5.0 4.0 3.0 2.0 1.0 0.0 Fast neutron fluence (×10 24 n/m 2 ) Distance from top of graphite reflector (mm) Fig.4 (b). Comparison of the cross-section average between the dimensional change and fast neutron fluence as a function of the distance from top of the graphite reflector (In the case that a part of the graphite reflector obviously pushed the aluminum cladding). 312 of 455
2 Demensional change (%) 1.5 1 0.5 0 0 1 2 3 4 5 6 Fast neutron fluence(×10 24 n/m 2 ) Fig.5. Dimensional change of the graphite under the JRR-4 operation condition as a function of the fast neutron fluence. 4. Conclusion We found that a large irradiation growth happened to the graphite reflectors under JRR-4 operation condition, of which temperature was lower than 200 ˚C. The survey on the reflector elements confirmed that the crack was caused by irradiation growth of the graphite material by fast neutrons. The irradiation growth ratio was 7.13x10 -25 %m 2 /n in the range of fast neutron fluence below 2.5x10 24 n/m 2 . We will take this result into account in the future reflector design. 5. References [1] H. Matsuo: “Effects of High Temperature Neutron Irradiation on Dimensional Changes and Physical Properties of <strong>Nuclear</strong> Graphites for HTGR”, JAERI-M 87-207 (1987) [in Japanese]. [2] H. Bridge, B.T. Kelly and P.T. Nettley: “Effect of High-Flux Fast-Neutron Irradiation on the Physical Properties of Graphite”, Carbon 2, 83-93 (1964). [3] K. Okumura, T. Kugo, K. Kaneko and K. Tsuchihashi: “SRAC2006: A Comprehensive Neutronics Calculation Code System”, JAEA-Data/Code 2007-004 (2007). [4] X-5 Monte Carlo Team: “MCNP-A General Monte Carlo N-Particle Transport Code, Version 5”, LA-UR-03-1987 (2003). 313 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 264 and 265: Page 7 / 7 References: [1] M.C. Ans
Page 266 and 267: Nuclear Research Institutes, and so
Page 268 and 269: According to the latest IAEA inform
Page 270 and 271: 18. May Training of operating perso
Page 272 and 273: 2.1 Irradiation Facilities Brief te
Page 274 and 275: 3.1 Detection limits Although it is
Page 276 and 277: PARTIAL DISMANTLING OF RESEARCH REA
Page 278 and 279: • Core - liable to full scale rep
Page 280 and 281: • Necessary individual dosimetry
Page 282 and 283: SILICON DOPING AT FRM II H. GERSTEN
Page 284 and 285: The resulting neutron flux density
Page 286 and 287: 4. References [1] Wagner F.M., Knes
Page 288 and 289: The choice of fuel base and solutio
Page 290 and 291: 3.3 Increasing power beyond current
Page 292 and 293: A COATING TO PROTECT SPENT ALUMINIU
Page 294 and 295: The treatment consisted of simple i
Page 296 and 297: 5. Conclusions 1. Immersion of AA 1
Page 298 and 299: KEEPING AGING RESEARCH REACTORS IN
Page 300 and 301: inspected in presence of an expert
Page 302 and 303: After the inspection was finished,
Page 304 and 305: 1 A STRATEGY FOR MANAGING AGEING CO
Page 306 and 307: 3 2. With the aluminium-clad HEU fu
Page 308 and 309: 5 The pool and the support structur
Page 310 and 311: Fig.2. Reflector element of JRR-4.
Page 314 and 315: SAFETY CONSIDERATIONS FOR CORE MANA
Page 316 and 317: 3 3. Core operation and monitoring
Page 318 and 319: 5 temperature, etc.). Effective sec
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Page 322 and 323: compared with that of the delayed n
Page 324 and 325: AD-HOC AND PREVENTATIVE MAINTENANCE
Page 326 and 327: Page 3 of 7 3. SAFARI-1OPERATIONAL
Page 328 and 329: Page 5 of 7 programs for instrument
Page 330 and 331: Page 7 of 7 DATE DESCRIPTION DATE D
Page 332 and 333: MICROSTRUCTURAL ANALYSIS OF MTR FUE
Page 334 and 335: In sample S3, similar zones could b
Page 336 and 337: increased temperature, the solid st
Page 338 and 339: Figure 8 BEI images covering Sample
Page 340 and 341: 660 °C, but since in the former me
Page 342 and 343: support. Both Governments have eval
Page 344 and 345: • Removal of sludge from the SNF
Page 346 and 347: UPGRADED REACTOR SYSTEMS FOR ENHANC
Page 348 and 349: Parameter HEU LEU a 6.0x10 -5 (°K)
Page 350 and 351: From 1992 until 2005 the TRIGA-SSR
Page 352 and 353: Figure9 - Dependency between reacto
Page 354 and 355: URANIUM CONTAMINATION IN PRIMARY CI
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Page 358 and 359: 6. Acknowledgement This work was pe
Page 360 and 361: 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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