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Tsukimori, K.; Kikuchi, M.; Iwata, K.
Transactions of the 11th international conference on structural mechanics in reactor technology1991
Transactions of the 11th international conference on structural mechanics in reactor technology1991
AbstractAbstract
[en] Main conclusions are as follows. (1) The buckling pressures can be evaluated by using the simple formula of in-plane instability. However, the formula based on the external pressure buckling of the equivalent cylinder gives an unconservative estimation. (2) The details of buckling behaviors (buckling pressure and buckling mode) can be evaluated numerically by using the load stiffness matrix. The numerical efficiency can be achieved by using the harmonic series shell element. (3) There is a possibility of unstable failure immediately after buckling. (4) The buckling pressure is almost independent of the number of convolutions. (author)
Primary Subject
Source
Shibata, Heki (ed.) (Tokyo Univ. (Japan). Inst. of Industrial Science); Atomic Energy Society of Japan, Tokyo (Japan); 6297 p; 1991; v. E p. 327-332; Atomic Energy Society of Japan; Tokyo (Japan); 11. international conference on structural mechanics in reactor technology; Tokyo (Japan); 18-23 Aug 1991
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Book
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Conference
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Hirano, K.; Ohmatsuzawa, K.; Takeda, T.; Nakane, S.; Kawaguchi, T.; Nagao, K.
International conference on fast reactors and related fuel cycles1991
International conference on fast reactors and related fuel cycles1991
AbstractAbstract
[en] Physical property tests of concrete subjected to high temperature were performed to obtain basic data to be reflected in design of nuclear reactor buildings for 'Monju'. Based on the result of tests on mechanical properties, thermal deformation properties, and thermal constants of concrete in a temperature range of 20 to 175degC, studies were made of the influences of various factors on the physical properties of concrete subjected to high temperature. (author)
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Atomic Energy Society of Japan, Tokyo (Japan); [2900 p.]; 1991; v. 3 p. 2.25/1-2.25/10; Atomic Energy Society of Japan; Tokyo (Japan); International conference on fast reactors and related fuel cycles; Kyoto (Japan); 28 Oct - 1 Nov 1991
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Book
Literature Type
Conference
Country of publication
BREEDER REACTORS, BUILDING MATERIALS, BUILDINGS, CONTAINMENT, EPITHERMAL REACTORS, EXPANSION, FAST REACTORS, FBR TYPE REACTORS, LIQUID METAL COOLED REACTORS, LMFBR TYPE REACTORS, MATERIALS, MECHANICAL PROPERTIES, MOISTURE, PHYSICAL PROPERTIES, POWER REACTORS, REACTORS, SODIUM COOLED REACTORS, TESTING, THERMODYNAMIC PROPERTIES
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INIS VolumeINIS Volume
INIS IssueINIS Issue
AbstractAbstract
[en] As for light water reactors, the structure is relatively simple, and the power plants of large capacity can be realized easily, therefore, they have been used for long period as main nuclear reactors. During that period, the accumulation of experiences on the design, manufacture, operation, maintenance and regulation of light water has become enormous, and in Japan, the social base for maintaining and developing light water reactor technologies has been prepared sufficiently. If the nuclear power generation using seawater uranium is considered, the utilization of uranium for light water reactor technologies can become the method of producing the own energy for Japan. As the factors that threaten the social base of light water reactor technologies, there are a the lowering of the desire to promote light water reactors, the effect of secular deterioration, the price rise of uranium resources, the effect of plutonium accumulation, the effect of the circumstances in developing countries and the sure recruiting of engineers. The construction and the principle of working of light water reactors and the development of light water reactors hereafter, for example, the improvement on small scale and the addition of new technology resulting in cost reduction and the lowering of the quality requirement for engineers, the improvement of core design, the countermeasures by design to serious accidents and others are described. (K.I.)
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Journal Article
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AbstractAbstract
[en] The ABAQUS finite element code was used to model a pressurized pipe and subjected to cyclic bending loads to investigate ratcheting. A 1-in. schedule 40 pipe was loaded with a slow (static) cyclic load. The pipe internal pressure was varied from 0 to 6000 psi. In this paper, two types of materials were considered: an elastic perfectly plastic and a bilinear elastic-plastic material. Two types of finite elements of the ABAQUS program were compared to analytical solutions to evaluate the element accuracy in the plastic regime. Depending upon loading conditions and specified material properties, three different responses were observed from the finite element analyses: cyclic plasticity, ratcheting of the hoop strain, or shakedown. These analytical results are compared to some experimental measurements
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Journal Article
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AbstractAbstract
[en] Flanges in piping systems and also pressure vessel flanges on tall columns are often subjected to longitudinal bending moments of considerable magnitude, be it from thermal expansion stresses in piping systems or from wind or seismic loadings on tall vertical pressure vessels. Except for the ASME Code, Section III, Subsections NB, NC, and ND, other pressure vessel and piping codes do not contain design ASME Nuclear Power Plant Code (Section III), an empirical formula is given, expressing a longitudinal bending moment in bolted flanged connections in terms of an equivalent internal pressure to be added to the design pressure of the flange. In this paper, an attempt is made to analyse the stresses on flanges and bolting due to external bending moments and to compare flange thicknesses thus obtained with thicknesses required using the equivalent design pressure specified in Subsections NB, NC, and ND. A design method is proposed, based on analysis and experimental work, which may be suitable for flange bending moment analysis when the rules of the Nuclear Power Plant Code are not mandatory. (orig.)
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Vereinigung der Technischen Ueberwachungsvereine e.V., Essen (Germany); 857 p; 1992; p. 81-89; 7. international conference on pressure vessel technology (ICPVT-7); 7. Internationale Konferenz ueber Druckbehaeltertechnologie; Duesseldorf (Germany); 31 May - 5 Jun 1992; Available from FIZ Karlsruhe
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Miscellaneous
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Conference
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AbstractAbstract
[en] The paper describes tests to determine the leakage behavior of inflatable seals when subjected to containment pressures that exceed the design basis. Inflatable seals are used to prevent leakage around personnel and escape lock doors in about 10% of the commercial nuclear power plant containment structures in the United States. All of the installations are in either Pressurized Water Reactor (PWR) or Boiling Water Reactor (BWR) Mark-III type containments. This work is a part of an overall effort at Sandia National Laboratories to develop proven techniques for evaluating the performance of Light Water Reactor (LWR) containment buildings for beyond design basis loadings. Inflatable seals were tested at both room temperature and at elevated temperatures representative of postulated severe accident conditions. Parameters that were monitored and recorded during each test were the internal seal pressure and temperature, chamber (containment) pressure, leakage past the seals, and temperature of the test chamber and fixture to which the seals were attached. An empirically based, analytical method is presented to predict the containment pressure at which significant leakage past inflatable seals can be expected. (orig.)
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CONTRACT DE-AC04-76DP00789
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Journal Article
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Hirano, K.; Ohmatsuzawa, K.; Takeda, T.; Ohkuma, M.; Akiyama, K.
International conference on fast reactors and related fuel cycles1991
International conference on fast reactors and related fuel cycles1991
AbstractAbstract
[en] The inner concrete structure of the FBR power station, 'Monju' is designed hypothesizing cases of the facility being subjected to high temperature at time of the sodium leakage accident, and of being subjected to earthquake while in that state. This report describes the technique applied to 'Monju' as a rational and practical design technique for a reinforced concrete structure subjected to such thermal load and combination of thermal load and seismic load. Based on this technique, the evaluation methods for thermal stress, for seismic load and seismic stress, and for member strengths are described. (author)
Primary Subject
Source
Atomic Energy Society of Japan, Tokyo (Japan); [2900 p.]; 1991; v. 3 p. 2.26/1-2.26/10; Atomic Energy Society of Japan; Tokyo (Japan); International conference on fast reactors and related fuel cycles; Kyoto (Japan); 28 Oct - 1 Nov 1991
Record Type
Book
Literature Type
Conference
Country of publication
ACCIDENTS, ALKALI METALS, BREEDER REACTORS, BUILDING MATERIALS, COMPOSITE MATERIALS, CONCRETES, ELEMENTS, EPITHERMAL REACTORS, FAST REACTORS, FBR TYPE REACTORS, LIQUID METAL COOLED REACTORS, LMFBR TYPE REACTORS, MATERIALS, MECHANICAL PROPERTIES, METALS, POWER REACTORS, REACTOR ACCIDENTS, REACTORS, REINFORCED MATERIALS, SODIUM COOLED REACTORS, STRESSES, TEMPERATURE RANGE
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Nagakura, Hiroshi; Kaneko, Shigehiko
Transactions of the 11th international conference on structural mechanics in reactor technology1991
Transactions of the 11th international conference on structural mechanics in reactor technology1991
AbstractAbstract
[en] The stability of a cantilever beam subjected to one-dimensional leakage flow is studied both theoretically and experimentally. It is clarified that in the case that the beam is clamped at the upstream end, the system loses stability by coupled-mode flutter, on the other hand, in the case that the beam is clamped at the downstream end, the system first loses stability by divergence and successively loses stability by flutter with increasing flow velocity. (author)
Primary Subject
Source
Shibata, Heki (ed.) (Tokyo Univ. (Japan). Inst. of Industrial Science); Atomic Energy Society of Japan, Tokyo (Japan); 6297 p; 1991; v. J p. 135-140; Atomic Energy Society of Japan; Tokyo (Japan); 11. international conference on structural mechanics in reactor technology; Tokyo (Japan); 18-23 Aug 1991
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Book
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Conference
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INIS VolumeINIS Volume
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Hayashi, Kimio; Amano, Hidetoshi; Suzuki, Yasufumi; Furuta, Teruo; Nagase, Fumihisa; Suzuki, Masahide
Japan Atomic Energy Research Inst., Tokyo (Japan)1993
Japan Atomic Energy Research Inst., Tokyo (Japan)1993
AbstractAbstract
[en] A concept of the Very High Burnup Fuel aiming at a maximum fuel assembly burnup of 100 GWd/t has been proposed in terms of burnup extension, utilization of Pu and transmutation of transuranium elements (TRU: Np, Am and Cm). The authors have investigated research and development (R and D) subjects of the fuel pellet and the cladding material of the Fuel. The present report describes the results on the fuel pellet. First, the chemical state of the Fuel and fission products (FP) was inferred through an FP-inventory and an equilibrium-thermodynamics calculations. Besides, knowledge obtained from post-irradiation examinations was surveyed. Next, an investigation was made on irradiation behavior of U/Pu mixed oxide (MOX) fuel with high enrichment of Pu, as well as on fission-gas release and swelling behavior of high burnup fuels. Reprocessibility of the Fuel, particularly solubility of the spent fuel, was also examined. As for the TRU-added fuel, material property data on TRU oxides were surveyed and summarized as a database. And the subjects on the production and the irradiation behavior were examined on the basis of experiences of MOX fuel production and TRU-added fuel irradiation. As a whole, the present study revealed the necessity of accumulating fundamental data and knowledge required for design and assessment of the fuel pellet, including the information on properties and irradiation performance of the TRU-added fuel. Finally, the R and D subjects were summarized, and a proposal was made on the way of development of the fuel pellet and cladding materials. (author)
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Jun 1993; 185 p
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Report
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Nagase, Fumihisa; Suzuki, Masahide; Furuta, Teruo; Suzuki, Yasufumi; Hayashi, Kimio; Amano, Hidetoshi
Japan Atomic Energy Research Inst., Tokyo (Japan)1993
Japan Atomic Energy Research Inst., Tokyo (Japan)1993
AbstractAbstract
[en] Plutonium use as well as burnup extension of UO2 fuel is an important subject for the strategy of utilization of the nuclear energy in LWRs. A higher burnup is favorable to MOX fuel in economic respect and for effective use of plutonium. Therefore, the concept of a 'very high burnup' aiming at the maximum bundle burnup of 100GWd/t has been proposed assuming use of MOX fuel. The authors have investigated research and development subjects for the fuel pellet and the cladding material to be developed. The present report shows the results on the cladding material. In order to achieve a very high burnup, development of the cladding material with higher corrosion and radiation resistance compared with Zircaloy is necessary. In this report, zirconium based alloy, stainless steel, nickel and titanium based alloys, ceramics, etc. were reviewed considering water corrosion resistance, thermal and mechanical properties, radiation effects, etc. Furthermore, capability of these materials as the fuel cladding was discussed focusing on water side corrosion and radiation effect on mechanical properties. As a result, candidate materials at present and the required research tasks were shown with issues for the development. (author) 66 refs
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May 1993; 124 p
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Report
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