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China Inst. of Atomic Energy, Beijing (China); 241 p; ISBN 7-5022-1329-5; ; 1995; p. 140-141
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China Inst. of Atomic Energy, Beijing (China); 101 p; ISBN 7-5022-2075-5; ; Oct 1999; p. 35
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AbstractAbstract
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China Inst. of Atomic Energy, Beijing (China); 241 p; ISBN 7-5022-1329-5; ; 1995; p. 120
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Book
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Numerical Data
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AbstractAbstract
[en] The FGRAT code was used for simulating calculation of fission gases axial transportation based on the data of Qinshan NPP test rods. Also the effect caused by fission gas local concentration on the fuel rod center temperature was evaluated. Some useful results were obtained
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China Inst. of Atomic Energy, Beijing (China); 232 p; ISBN 7-5027-2965-8; ; 1992; p. 149; China Ocean Press; Beijing (China)
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Book
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CALCULATION METHODS, COMPUTER CODES, ELEMENTS, ENRICHED URANIUM REACTORS, FUEL ELEMENTS, MATHEMATICAL MODELS, NONMETALS, NUCLEAR FACILITIES, NUMERICAL SOLUTION, POWER PLANTS, POWER REACTORS, PWR TYPE REACTORS, RARE GASES, REACTOR COMPONENTS, REACTORS, THERMAL POWER PLANTS, THERMAL REACTORS, WATER COOLED REACTORS, WATER MODERATED REACTORS
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AbstractAbstract
No abstract available
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China Inst. of Atomic Energy, Beijing (China); 311 p; ISBN 7-5027-2171-1; ; 1991; p. 133; China Ocean Press; Beijing (China)
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Book
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Zhang Yingchao
Water reactor fuel element computer modelling in steady state, transient and accident conditions1989
Water reactor fuel element computer modelling in steady state, transient and accident conditions1989
AbstractAbstract
[en] Recently at the Institute of Atomic Energy (IAE), RELAP5/MOD1, FRAPCON-2, and FRAP-T6 have been linked for analyzing the fuel behavior under Small-Break Loss of Coolant Accident (SB-LOCA) conditions for Qinshan Nuclear Power Plant (Qinshan NPP) which is under construction in China. The main results are quite reasonable, but some problems have been found in combining the functions of different codes, for example, when using the initial conditions provided by FRAPCON-2, FRAP-T6 gives incredibly large cladding stress at the beginning of the accident. It is suggested that for these codes some further improvements are desirable. (author). 4 refs, 8 figs, 2 tabs
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International Atomic Energy Agency, Vienna (Austria). International Working Group on Water Reactor Fuel Performance and Technology; 322 p; May 1989; p. 234-241; Technical committee on water reactor fuel element computer modelling in steady state, transient and accident conditions; Preston (UK); 18-22 Sep 1988; IAEA-TC--659/4.6
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Report
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AbstractAbstract
[en] The annular fuel that cooled internally and externally is a new PWR fuel which consists of two cladding and annular pellets. The annular fuel has more safety performance for new structure. Qinshan Phase Ⅱ Nuclear Power Plant project was selected to analyze large break loss of coolant accident (LBLOCA) as the reference plant. The results show that the plant loaded with the annular fuel is more safe than that loaded with the solid fuel during LBLOCA. (authors)
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3 figs., 2 tabs., 5 refs.
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Journal Article
Journal
Atomic Energy Science and Technology; ISSN 1000-6931; ; v. 48(suppl.); p. 463-466
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AbstractAbstract
[en] The calculational model of ICARE2 V2mode2.3 has been built for Qinshan Nuclear Power Plant and the simulation of early core degradation of SBLOCA with station blackout has been performed. The calculation results show that the core starts damage from 4560s and the whole process of early core degradation is relatively fast due to poor core cooling conditions. The analysis of early core degradation is only valid up to around 7900s
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Journal Article
Journal
Atomic Energy Science and Technology; ISSN 1000-6931; ; v. 34(suppl.); p. 82-85
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AbstractAbstract
[en] The Japanese fuel behavior analysis code FEMAXI-IV provided by OECD is used to calculate the stress level of Qinshan Nuclear Power Plant fuel rod under pellet clad mechanical interaction (PCMI) and to evaluate the safety margin of PCMI. The threshold of rod power is obtained after the determination of threshold of local hoop stress. The safety margin of linear power is 147 W/cm and the safety margin of local hoop stress is 263 MPa under the accident condition of unanticipated control rod withdrawing according to the threshold
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Journal Article
Journal
Atomic Energy Science and Technology; ISSN 1000-6931; ; v. 34(suppl.); p. 86-89
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AbstractAbstract
[en] Using the measured leakage rate of the CARR building, MELCOR code was used to calculate the pressure of the building under the severe accident conditions. The results indicate that the maximum pressure difference in the progress of the accident is less than 5 kPa which is much lower than the structure design limiting value of 10 kPa; the calculated maximum leakage rates are lower than vol.2.5%/d in 8 hours and vol.2.8%/d in l day after the initiation of the accident. The dose of the radioactive fission product released to the environment is lower than the safety criteria under the leakage rate. (authors)
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7 figs., 2 refs.
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Journal Article
Journal
Atomic Energy Science and Technology; ISSN 1000-6931; ; v. 44(suppl.); p. 258-260
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