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Design, construction and operation of nuclear power plants conference; Portland, OR (USA); 5-8 Aug 1984; CONF-840813--; Published in summary form only.
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Transactions of the American Nuclear Society; ISSN 0003-018X;
; v. 46(1); p. 104-105
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Joint meeting of the American Nuclear Society and the Atomic Industrial Forum; Washington, DC (USA); 11-16 Nov 1984; CONF-841105--; Published in summary form only.
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Transactions of the American Nuclear Society; ISSN 0003-018X;
; v. 47 p. 527-529
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ANS international conference; Washington, DC (USA); 17 - 21 Nov 1980; CONF-801107--; Published in summary form only.
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Transactions of the American Nuclear Society; ISSN 0003-018X;
; v. 35 p. 419
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American Nuclear Society winter meeting; San Francisco, CA (USA); 30 Oct - 4 Nov 1983; CONF-831047--; Published in summary form only.
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Transactions of the American Nuclear Society; ISSN 0003-018X;
; v. 45 p. 393-394
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American Nuclear Society's annual meeting; Miami Beach, FL (USA); 7 - 12 Jun 1981; CONF-810606--; Published in summary form only.
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Transactions of the American Nuclear Society; ISSN 0003-018X;
; v. 38 p. 299-300
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AbstractAbstract
[en] Silicon carbide (SiC) based composite is considered among leading candidate materials for accident tolerant fuel (ATF) cladding in light water reactors (LWRs) because of its outstanding irradiation and oxidation resistance, as well as low neutron absorption. Under normal operation conditions, the synergism between neutron irradiation and a high radial heat flux though the cladding thickness is expected to produce significant stress. This stress is attributed to a significant through-thickness temperature gradient and the resulting differential swelling strain due the highly temperature-dependent point-defect swelling of SiC. Although modeling of the in-pile stress state of SiC cladding has been conducted using finite element analysis, the analysis is challenging because of the complex stress state that depends upon swelling, irradiation creep, and thermal conductivity, all of which are both temperature and dose-dependent. These integral models have not yet been verified by experiments. In order to experimentally validate the multi-physics thermo-mechanical models of ATF SiC cladding during operation, SiC tube specimens have been irradiated under a high radial heat flux. This paper reports preliminary results from the post-irradiation examination (PIE) in on-going project supported by the U.S. Department of Energy's Nuclear Science User Facility program. The objectives of the PIE are (1) determination of the irradiation temperature distribution within the tube specimens, and (2) evaluation of the state and magnitude of the stress within the tube specimens arising from the temperature gradient. This paper mainly presents findings for objective (1)
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2017 Annual Meeting of the American Nuclear Society; San Francisco, CA (United States); 11-15 Jun 2017; Country of input: France; 7 refs.; available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US)
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Transactions of the American Nuclear Society; ISSN 0003-018X;
; v. 116; p. 386-388
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ABSORPTION, ACCIDENT-TOLERANT NUCLEAR FUELS, CLADDING, CREEP, FINITE ELEMENT METHOD, HEAT FLUX, IRRADIATION, NEUTRONS, OXIDATION, POINT DEFECTS, POST-IRRADIATION EXAMINATION, SILICON CARBIDES, STEADY-STATE CONDITIONS, SWELLING, TEMPERATURE DEPENDENCE, TEMPERATURE GRADIENTS, THERMAL CONDUCTIVITY, TUBES, WATER COOLED REACTORS, WATER MODERATED REACTORS
BARYONS, CALCULATION METHODS, CARBIDES, CARBON COMPOUNDS, CHEMICAL REACTIONS, CRYSTAL DEFECTS, CRYSTAL STRUCTURE, DEFORMATION, DEPOSITION, ELEMENTARY PARTICLES, ENERGY SOURCES, FERMIONS, FUELS, HADRONS, MATERIALS, MATHEMATICAL SOLUTIONS, MECHANICAL PROPERTIES, NUCLEAR FUELS, NUCLEONS, NUMERICAL SOLUTION, PHYSICAL PROPERTIES, REACTOR MATERIALS, REACTORS, SILICON COMPOUNDS, SORPTION, SURFACE COATING, THERMODYNAMIC PROPERTIES
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[en] Two situations arose recently in which the determination of component service life was important. In both situations, irradiation in a research reactor and analytic guidance and interpretation of results provided the service-life information. The first situation involved cables for postaccident neutron detectors subjected to both neutron and gamma irradiation. As described further, cable specimens were irradiated in the neutron and gamma environment of a research reactor. The results were compared with calculational models of the power reactor to show environmental qualification for the life of the plant. The second situation subjected replacement neutron detectors to separate neutron and gamma irradiations in order to provide service-life information
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Winter meeting of the American Nuclear Society (ANS) and the European Nuclear Society (ENS); Washington, DC (United States); 10-14 Nov 1996; CONF-961103--
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[en] Leibstadt the first 1000 MW class BWR plant in Switzerland was subject to a comprehensive range of tests. The unit included some newly designed equipment on which special tests were made. The progress of the commissioning was subjected to safety authority permits, issued stepwise upon successful conclusion of previous checks. The zone within which the reactor could operate without instabilities was established, and the proper functioning of safety systems consequent upon the introduction of fault conditions was demonstrated. The test results confirmed the values expected and the plant's efficiency and output figures were better than guaranteed. The fulfillment of the commissioning program has, in part, contributed to Leibstadt's first class early commercial performance
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6. Pacific Basin nuclear conference; Beijing (China); 7-11 Sep 1987; CONF-870905--
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[en] The nuclear industry's shift toward total quality management, renewed emphasis on stating requirements and achieving consistent performance, is the subject of this paper. The impact of total quality management centers on three outward-focused areas: refined skills, performance, and process management. Process management requires a total change in organizational philosophy, policies, and management practice
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Annual meeting of the American Nuclear Society (ANS); Orlando, FL (United States); 2-6 Jun 1991; CONF-910603--
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[en] The objective of the subject work was to develop and evaluate conservative but accurate homogeneous model presciptions for heat transfer in and among heterogeneous modular high-temperature gas-cooled reactor (MHTGR) fuel blocks. Such simplifications are useful in analyses of complex transients where it is impractical to describe the core blocks in precise detail
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American Nuclear Society annual meeting; Boston, MA (United States); 7-12 Jun 1992; CONF-920606--
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