RRFM 2009 Transactions - European Nuclear Society

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Heat capacity Information Thermal conductivity Thermal expansion coefficient Exothermic energy release Mechanical properties Rationale for Providing this Information Needed to calculate the fuel meat temperature in nonequilibrium situations and needed for measurement of fuel meat thermal conductivity by the laser-flash method Needed to calculate fuel meat temperature during irradiation and during post-irradiation cooling Needed to evaluate stress resulting from differential thermal expansion during design of fuel element Needed to evaluate fuel temperatures during slow transients or loss-of-coolant situations Needed during design of the fuel element and fuel assembly Fuel Meat and Fuel Element Irradiation Properties Fission density distribution Fuel element and fuel meat swelling Fuel meat and fuel particle microstructures Mechanical integrity Blister threshold temperature Cladding corrosion behaviour Fission product release Needed itself and needed to estimate the time history of fuel fission rates, both of which are required to interpret data from post-irradiation examinations Needed to determine if fuel is behaving in a stable manner and to estimate the narrowing of coolant channels between fuel elements Needed to accurately characterize the nature of fuel/matrix or fuel/cladding interaction and the nature of fuel material and fuel meat swelling Needed to assure that an adequate heat transfer path and containment of fission products is maintained Provides a figure of merit of the fuel element’s ability to withstand high temperatures during incidents Needed to assure integrity against the release of fission products and to estimate the thermal history of the fuel element Needed to estimate the radiological source term in accident situations Fuel Assembly Properties Hydraulic and mechanical properties Needed to calculate coolant flow rates and to assure that the fuel assembly can withstand hydraulic and thermal stresses during irradiation Irradiation behaviour Needed to demonstrate that fuel assemblies will perform under irradiation as expected, based on all of the properties listed above in this table. Table 2. Information that should be provided by the fuel developer 120 of 455
2.5. Generic approach to fuel development and qualification A generic approach for developing and qualifying of a new fuel is described in the first three sections of Chapter 5. The fourth section describes the variations in the process needed when qualifying an already-qualified fuel type for use under the existing qualification limits (that is, when a potential user of a previously qualified fuel wants to use that fuel in a fuel element or fuel assembly with a different design than the one in which the fuel was originally qualified) and for use outside the existing qualification limits. The approach described is a consistent and integrated approach based on the experience of organizations previously involved in LEU fuel development and on international best practices. The description of the fuel development and qualification process focuses on the tasks to be accomplished and the logical sequence of events. The tasks are not described in detail and the way in which the tasks should be accomplished are not prescribed, because such details depend on the policies, practices, methodologies, and standards of the particular country and organizations involved. It is noted that before its implementation, the fuel qualification programme for a new fuel should be discussed with the Regulatory Body in order to address the requirements for additional data related to fuel behaviour under accident conditions. A phased approach is recommended for LEU fuel development and qualification, with the first phase focused on research work and development/testing activities leading to the selection of a preferred fuel type, the second phase focused on fuel performance qualification to demonstrate that the selected fuel meets specified performance requirements, and the third phase focused on qualification of the fuel manufacturer. The recommended activities are summarized in Table 3. Activity Phase 1: Research and Development Fuel conceptual design Fuel manufacturing development Out-of-pile testing Irradiation testing and postirradiation examination Results of Activity Rationale for developing a new fuel Choice of design concept for the fuel element and fuel assembly Initial selection of candidate fuel and cladding materials Initial selection of candidate manufacturing technologies Fissile material (meat) manufacturing process Cladding manufacturing process Fuel element (plates, rods, tubes) manufacturing process Basic fuel properties (before and after manufacturing) Compatibility of fuel meat components or fuel and cladding (before and after manufacturing) Integrity of the fuel meat/cladding bond Resistance of the fuel to corrosion in the presence of the coolant (after manufacturing) Determination of required materials properties and performance levels Selection of irradiation devices and facilities Preparation of safety report Performance of in-pile measurements 121 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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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 96 and 97: Finally, a new tool was defined and
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 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
Page 146 and 147: uniform Si diffusion and consequent
Page 148 and 149: Recently, fuel relocation induced b
Page 150 and 151: characteristics of ILs obtained in
Page 152 and 153: - Type 2 (Si content ≥ 5 wt%) : c
Page 154 and 155: 4. Discussion The main contribution
Page 156 and 157: METHODS OF INCREASING THE URANIUM C
Page 158 and 159: • strength properties of Zr-1% Nb
Page 160 and 161: 4. References 1. Birzhevoy G.A., Ka
Page 162 and 163: Dispersion fuel Monolithic fuel ATR
Page 164 and 165: − because of its Newtonian charac
Page 166 and 167: Session III Back-end 166 of 455
Page 168 and 169: 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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RRFM 2009 Transactions - European Nuclear Society

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