RRFM 2009 Transactions - European Nuclear Society

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Finally, a new tool was defined and the validation was obtained through several pull tests performed on dummy sections representative of the element and thorough inspections of each watergap – see example of inspection with capacitive gauges on figure 10. You can notice that results are as homogeneous from the most internal to the most external watergap. watergap 2,50 2,40 2,30 UPPER LIMIT 2,20 2,10 2,00 1,90 1,80 1,70 1,60 1,50 1 32 63 94 125 156 187 218 249 280 311 342 373 404 435 466 497 528 559 mm 590 621 652 683 LOWER LIMIT measurement points Figure 10 – Inspection of the watergaps on JHR fuel section using capacitive gauges 3.4. Fabrication of an assembly able to be easily dismantled For the elements meant for EOLE installation, one requirement was to benefit from an element able to be easily dismantled in order to measure each plate by gamma-scanning and to have more precise information about neutron beam. The challenge for AREVA- CERCA was to design a fuel element almost identical to the others, but that could be completely dismantled. The main requirements of the end-user were the following: - Same watergap (1,95 mm) with a tolerance enlarged from 0,30 mm to 0,50 mm, - Precise positioning of plates in the element and maintaining of their position, - Assembling and dismantling of the element at least 100 times with a facilitated extraction of the plates. Advanced studies were performed during a period of 10 months to find adequate solutions to each of these specificities. The work consisted in determining a set of solutions for each requirement, presenting the principles to the end-user to take into account its constraints (a few materials prohibited like stainless steel in different locations of the element, obstruction of watergaps as reduced as possible, …) and testing the retained solutions on an inert dummy at scale 1 to ensure of their adequacy with the given requirements. At least, a specific design of element was proposed to the CEA, integrating adapted solutions for each of the specificities requested: for example, specific lateral comb enabling to maintain the watergap as well as to ensure the precise positioning of plates in the fuel assembly, surface treated stiffeners to facilitate the extraction of plates, design of a made to measure assembling tool and so on. 96 of 455
Figures 11 and 12 show local views of a standard fuel assembly (figure 11) and of the assembly able to be easily dismantled for EOLE (figure 12). Surface treated stiffener Lateral comb Figure 11 – Sketch of a standard fuel assembly Figure 12 – Sketch of the fuel assembly able to be easily dismantled 4. Project schedule LTA's manufacturing for JHR fuel qualification program schedule 2006 2007 2008 <strong>2009</strong> Oct Nov Dec Jan Feb Mar Apr May June July Aug Sept Oct Nov Dec Jan Feb Mar Apr May June July Aug Sept Oct Nov Dec Jan Feb Mar Apr May June Contract between CEA and CERCA ♦ Preliminary study CEA / CERCA / AREVA TA meetings for project review ♦ ♦ ♦ ♦ ♦ ♦ ♦ poison plate qualification Fuel plate qualification technical specification validation ♦ ♦ ♦ structural parts supplying poison plate manufacturing Fuel plate manufacturing Fuel element manufacturing 1st EVITA fuel element witness inspection ♦ 4 EVITA fuel elements witness inspection ♦ 7 AMMON fuel elements witness inspection ♦ The project is still running; the first EVITA fuel element has been delivered in due time to BR2 reactor beginning of January <strong>2009</strong>, next step is the witness inspection for the last 4 EVITA fuel elements scheduled in March. This project schedule shows the last phase of JHR fuel qualification program, relative to the fuel plate manufacturing developments; but, previously to this phase, technical and design evaluation developments were successfully carried out, through a close collaboration between: the designer (AREVA-TA), the manufacturer (AREVA-CERCA) and 97 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
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
Page 94 and 95: A 3.1. Place a boron insert in a hi
Page 98 and 99: the end user (CEA), to define the f
Page 100 and 101: Previous papers discussed character
Page 102 and 103: (a) (b) (c) (d) (e) (f) (g) (h) (i)
Page 104 and 105: The results of a third reported irr
Page 106 and 107: In order to reproduce the heat tran
Page 108 and 109: 25 OXIDE THICKNESS 20 Kim et al pH=
Page 110 and 111: Heavy ion irradiation of UMo7/Al fu
Page 112 and 113: For the lowest flux values tested d
Page 114 and 115: 127 I beam 127 I beam 0 5 10 15 20
Page 116 and 117: Weight fractions (%) U(α) UO 2 γU
Page 118 and 119: activities to meet the need. The co
Page 120 and 121: Heat capacity Information Thermal c
Page 122 and 123: Decision point Activity Phase 2: Fu
Page 124 and 125: IRIS PROGRAM: IRIS4 FIRST RESULTS F
Page 126 and 127: The main characteristics of the IRI
Page 128 and 129: 4. In-pool plate thickness measurem
Page 130 and 131: At that stage of IRIS4 irradiation
Page 132 and 133: extrapolated thermal property value
Page 134 and 135: optical microscope, acoustic emissi
Page 136 and 137: CD WIRES AS BURNABLE POISON FOR THE
Page 138 and 139: Figure 1. MCNP geometry model of fu
Page 140 and 141: considered carefully: (i) maintain
Page 142 and 143: discussed and validated together, k
Page 144 and 145: 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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