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Rolstad, E.; Knudsen, K.D.; Hancvik, A.; Svanholm, K.
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
AbstractAbstract
[en] This paper describes two experiments, one dealing with clad overstraining failures and the other with fuel densification. The two series of experiments (IFA-229 overpower experiment on high burn-up fuel rods, and IFA-401 fuel densification experiment) have demonstrated the existence of two different fuel performance problems. When viewed individually, it appears rather simple to avoid both. However, they are associated with two opposite effects, the overstraining failures to fuel expansion and the fuel cladding collapses to fuel contraction. It is natural therefore that design changes introduced to avoid one of the problems will result in an adverse effect relative to the other
Primary Subject
Source
Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Committee on the safety of nuclear installations - CSNI, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France); Dept de Surete Nucleaire, Commissariat a l'Energie Atomique, CEN Saclay, 91191 Gif-sur-Yvette Cedex (France); 708 p; 1973; p. 223-237; Specialist meeting on the safety of water reactors fuel elements; Reunion de specialistes sur la surete des elements combustibles des reacteurs a eau; Saclay (France); 22-24 Oct 1973; 1 refs.
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Delafosse, J.; Mansard, B.
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
AbstractAbstract
[en] More or less adhesive deposits covering irradiated fuel elements in PWRs may, in some cases, interfere with the fuel-water thermal exchanges. This paper focuses on the cladding perforation mechanism that may result from these adhesive deposits in zones where thermal flux is high and flowrate is low. The nature and origin of the deposits has been analyzed; they are mainly composed of metal oxides derived from the reactor's primary circuit. Spinels (mixed iron and nickel oxides) are predominant; their magnetic properties could be used to remove them, since very thick deposits (several hundreds of microns) may derive from a pure water (15 to 30 iron ppb)
Original Title
Quelques resultats sur les depots formes en service sur les gaines d'elements combustibles
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Secondary Subject
Source
Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Committee on the safety of nuclear installations - CSNI, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France); Dept de Surete Nucleaire, Commissariat a l'Energie Atomique, CEN Saclay, 91191 Gif-sur-Yvette Cedex (France); 708 p; 1973; p. 39-66; Specialist meeting on the safety of water reactors fuel elements; Reunion de specialistes sur la surete des elements combustibles des reacteurs a eau; Saclay (France); 22-24 Oct 1973; 3 refs.
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CHEMICAL REACTIONS, COOLING SYSTEMS, CORROSION, DEPOSITION, ELEMENTS, ENERGY SYSTEMS, ENRICHED URANIUM REACTORS, FAILURES, FUEL ELEMENTS, HYDROGEN COMPOUNDS, INTERNATIONAL ORGANIZATIONS, METALS, MINERALS, OECD, OXIDE MINERALS, OXYGEN COMPOUNDS, POWER REACTORS, REACTOR COMPONENTS, REACTORS, SAFETY, SURFACE COATING, THERMAL REACTORS, TRANSITION ELEMENTS, WATER COOLED REACTORS, WATER MODERATED REACTORS
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AbstractAbstract
[en] Generally the problem of pellet densification and clad collapse in case of non fueled cladding have been attacked fervently since the discovery of fuel misbehaviour in a number of pressurized water reactors. However the causes for axial relocation are not made part of these studies in general. In this paper several questions relating to the mechanism of axial relocation are discussed, like means and modes of reactivity control as well as steady state operation. Design changes are reviewed in respect of the axial relocation of pellets, and possible mechanisms are presented
Primary Subject
Source
Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Committee on the safety of nuclear installations - CSNI, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France); Dept de Surete Nucleaire, Commissariat a l'Energie Atomique, CEN Saclay, 91191 Gif-sur-Yvette Cedex (France); 708 p; 1973; p. 103-122; Specialist meeting on the safety of water reactors fuel elements; Reunion de specialistes sur la surete des elements combustibles des reacteurs a eau; Saclay (France); 22-24 Oct 1973
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Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Committee on the safety of nuclear installations - CSNI, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France); Dept de Surete Nucleaire, Commissariat a l'Energie Atomique, CEN Saclay, 91191 Gif-sur-Yvette Cedex (France)
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
AbstractAbstract
[en] Fuel from the 440 MWe Oskarshamn I reactor, irradiated up to about 6000 MWd/t, has been examined by various methods with respect to fuel densification and axial gaps in the fuel stack. The conclusions are: fuel densification has occurred to a low but significant extent as measured on fuel stack length. The measured densification is within the extent suggested by the USAEC. The fuel stack length decrease is partly caused by pellet-pellet wear and adjustment. There exists a mechanism for pellet axial hang-up in the rods at thermal contraction at reactor shutdown. There were no indications of axial gaps in operating conditions. The gaps detected after reactor shutdown are filled out at power operation. Based on the results of the examination it can be statistically shown that the remainder of the core can not contain any large fraction of rods with gaps during operation and with a high level of confidence not coinciding gaps in adjacent rods
Primary Subject
Source
Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Committee on the safety of nuclear installations - CSNI, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France); Dept de Surete Nucleaire, Commissariat a l'Energie Atomique, CEN Saclay, 91191 Gif-sur-Yvette Cedex (France); 708 p; 1973; p. 139-158; Specialist meeting on the safety of water reactors fuel elements; Reunion de specialistes sur la surete des elements combustibles des reacteurs a eau; Saclay (France); 22-24 Oct 1973; 6 refs.
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Report
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ACTINIDE COMPOUNDS, BWR TYPE REACTORS, CHALCOGENIDES, DEPOSITION, ENRICHED URANIUM REACTORS, INTERNATIONAL ORGANIZATIONS, OECD, OPERATION, OXIDES, OXYGEN COMPOUNDS, PELLETS, POWER REACTORS, REACTOR COMPONENTS, REACTORS, SAFETY, SHUTDOWN, SURFACE COATING, THERMAL REACTORS, URANIUM COMPOUNDS, URANIUM OXIDES, WATER COOLED REACTORS, WATER MODERATED REACTORS
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Marstrand, J.
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
AbstractAbstract
[en] The author discusses the idea of improving safety by using fuel elements with deflectors for steam water separation in boiling water reactors. He recalls experiences carried out by a SNECMA-AEG project to establish rotation outside the rods by using twisted tapes (the project was abandoned), and considers various possibilities of flows for raising the burn-out limits such as using bigger subchannels (obtained by omitting rods in the regular networks) and other arrangements with and without deflectors
Primary Subject
Source
Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Committee on the safety of nuclear installations - CSNI, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France); Dept de Surete Nucleaire, Commissariat a l'Energie Atomique, CEN Saclay, 91191 Gif-sur-Yvette Cedex (France); 708 p; 1973; p. 397-410; Specialist meeting on the safety of water reactors fuel elements; Reunion de specialistes sur la surete des elements combustibles des reacteurs a eau; Saclay (France); 22-24 Oct 1973; 4 refs.
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Karwat, H.; Steinhoff, F.; Agemar, F.; Hicken, E.
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
AbstractAbstract
[en] The main objectives of the Emergency Cooling Program (ECP) are to determine heat transfer coefficients under blowdown conditions, by measuring the power input, the temperature along the test section, the pressure and pressure drop behaviour as well as the fluid conditions and fluid flow mechanism. The test results are used to check and improve present design codes. This paper is concerned with the second part of the ECP tests, namely tests with a four-rod bundle of BWR-geometry. The test facility, instrumentation and procedure are described, and the experimental results of the four-rod bundle experiments are given. Experiments carried out in the test facility were then simulated using the blowdown simulation code BRUCH-S-E1. The reasons for the deviations with experimental results are discussed
Primary Subject
Source
Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Committee on the safety of nuclear installations - CSNI, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France); Dept de Surete Nucleaire, Commissariat a l'Energie Atomique, CEN Saclay, 91191 Gif-sur-Yvette Cedex (France); 708 p; 1973; p. 599-658; Specialist meeting on the safety of water reactors fuel elements; Reunion de specialistes sur la surete des elements combustibles des reacteurs a eau; Saclay (France); 22-24 Oct 1973; 24 refs.
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Knudsen, P.; Hagen, H.H.; Stiff, J.
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
AbstractAbstract
[en] Fuel pins experience large and sometimes rapid variations in heat load during their life. The changes result from fuel element shuffling and changes in control rod positions. Increasing the pin power above a level which has been maintained for a long period may lead to failure because of fuel clad interaction. This type of failure will be dependent upon such factors as burnup and power level, rate and size of power increase, mode of operation and variables in the design of the pin. In order to quantify some of these relationships, well-characterized UO2-Zr fuel pins were irradiated to more than 20,000 MWD/t UO2 and then given an overpower ramp in special rigs. This report describes the experimental conditions and presents results obtained during the irradiation and post-irradiation examination. In the first test, a small amount of fission gas activity was observed but this was the result of previous surface contamination rather than cladding failure. In the second test, a failure was observed at a large ridge on the pin, and large local cladding deformations; it was concluded that the combination of heat load level and small gap resulted in excessive fuel clad interaction at the 20 pc overpower, which caused failure of the cladding
Primary Subject
Source
Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Committee on the safety of nuclear installations - CSNI, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France); Dept de Surete Nucleaire, Commissariat a l'Energie Atomique, CEN Saclay, 91191 Gif-sur-Yvette Cedex (France); 708 p; 1973; p. 91-102; Specialist meeting on the safety of water reactors fuel elements; Reunion de specialistes sur la surete des elements combustibles des reacteurs a eau; Saclay (France); 22-24 Oct 1973; 5 refs.
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ACCIDENTS, ACTINIDE COMPOUNDS, ALLOYS, CHALCOGENIDES, DEPOSITION, FAILURES, FUEL ELEMENTS, INTERNATIONAL ORGANIZATIONS, OECD, OXIDES, OXYGEN COMPOUNDS, REACTOR ACCIDENTS, REACTOR COMPONENTS, SAFETY, SURFACE COATING, SYSTEMS ANALYSIS, TESTING, TRANSITION ELEMENT ALLOYS, URANIUM COMPOUNDS, URANIUM OXIDES, ZIRCONIUM ALLOYS, ZIRCONIUM BASE ALLOYS
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Taketani, K.; Ichikawa, M.
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
AbstractAbstract
[en] The approach of JAERI's study on fuel safety for the investigation of the fuel failure mechanisms are as follows: 1. Pellet-cladding mechanical interaction at high heat rating (a cause of failure when the cladding becomes deteriorated by irradiation). 2. Cladding deterioration by hydriding and irradiation (the irradiation effect on the mechanical properties tends to saturate with neutron fluence, therefore it was postulated that the hydrogen addition to the cladding and irradiation of intermediate period of time will contribute to get the information on the high burnup state of the cladding). 3. Fuel failure by moisture in the fuel (considered as one of the major mechanisms of failure in the early life of the fuel): moisture was intentionally added to the fuel and the pin pressure was measured to study the rate of absorption of moisture)
Primary Subject
Source
Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Committee on the safety of nuclear installations - CSNI, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France); Dept de Surete Nucleaire, Commissariat a l'Energie Atomique, CEN Saclay, 91191 Gif-sur-Yvette Cedex (France); 708 p; 1973; p. 195-221; Specialist meeting on the safety of water reactors fuel elements; Reunion de specialistes sur la surete des elements combustibles des reacteurs a eau; Saclay (France); 22-24 Oct 1973; 17 refs.
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ASIA, BURNUP, CHEMICAL REACTIONS, DEPOSITION, DEVELOPED COUNTRIES, ELEMENTS, ENRICHED URANIUM REACTORS, INTERNATIONAL ORGANIZATIONS, NONMETALS, OECD, OPERATION, PELLETS, POWER REACTORS, REACTOR COMPONENTS, REACTORS, SAFETY, SURFACE COATING, THERMAL REACTORS, WATER COOLED REACTORS, WATER MODERATED REACTORS
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AbstractAbstract
[en] The design limits for the permissible discharge of radioactive gaseous effluents from nuclear stations in normal operation may be subject to variation from time to time. In the light of this there is a need for continuing re-assessment of the assumptions and data used in calculating the possible magnitude of gaseous effluents from water reactors to be built in the future. The interpretation placed on the available information reflects back on the integrity likely to be demanded from the fuel elements. Important aspects include the release of volatile fission products from failed fuel, the distribution of the fission products between water, steam, surfaces, and the off-gases, the possibilities for filtration and deposition outside the reactor, and the form of the species eventually released to atmosphere. This paper discusses some of these topics, making use of experience gained within the UKAEA's prototype Steam Generating Heavy Water Reactor (SGHWR) and comparing and contrasting this with information gained elsewhere where this is helpful. Some areas where information is lacking are noted and possible advantageous effects which could merit investigation are speculatively identified
Primary Subject
Source
Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Committee on the safety of nuclear installations - CSNI, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France); Dept de Surete Nucleaire, Commissariat a l'Energie Atomique, CEN Saclay, 91191 Gif-sur-Yvette Cedex (France); 708 p; 1973; p. 283-294; Specialist meeting on the safety of water reactors fuel elements; Reunion de specialistes sur la surete des elements combustibles des reacteurs a eau; Saclay (France); 22-24 Oct 1973; 15 refs.
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ELEMENTS, ENRICHED URANIUM REACTORS, FLUIDS, FUEL ELEMENTS, GASES, HALOGENS, HEAVY WATER MODERATED REACTORS, INTERNATIONAL ORGANIZATIONS, MATERIALS, NATIONAL ORGANIZATIONS, NONMETALS, OECD, POWER REACTORS, PRESSURE TUBE REACTORS, RADIOACTIVE MATERIALS, RADIOACTIVE WASTES, REACTOR COMPONENTS, REACTORS, SAFETY, THERMAL REACTORS, UNITED KINGDOM ORGANIZATIONS, WASTES, WATER COOLED REACTORS
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Bentley, M.J.; Trowse, F.W.
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
Proceedings of the specialist meeting on the safety of water reactors fuel elements1973
AbstractAbstract
[en] Experiments has shown that the ambient temperature embrittlement of Zircaloy fuel cladding subjected to a LOCA transient is principally due to embrittlement of the transformed beta phase by oxygen uptake. The bore hardness of the tube is a convenient indication of its brittleness. Hardness greater than 275 VPHN indicate failure at less than 10 pc diametral compression, and less than 225 VPHN at more than 20 pc compression. Use of the thickness ratio of retained alpha to transformed beta can be misleading as an indication of brittleness
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Source
Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Committee on the safety of nuclear installations - CSNI, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France); Dept de Surete Nucleaire, Commissariat a l'Energie Atomique, CEN Saclay, 91191 Gif-sur-Yvette Cedex (France); 708 p; 1973; p. 441-447; Specialist meeting on the safety of water reactors fuel elements; Reunion de specialistes sur la surete des elements combustibles des reacteurs a eau; Saclay (France); 22-24 Oct 1973; 3 refs.
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