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Zukeran, Atsushi*; Nakagawa, Tsuneo; Shibata, Keiichi; Ishikawa, Makoto*; Hino, Tetsushi*
JAERI-Research 2004-026, 102 Pages, 2005/02
JAERI-Research-2004-026.pdf:7.68MB
Reactivity uncertainties such as effective multiplication factor can be estimated by the sensitivity coefficients of the infinitely diluted cross section and resonance self-shielding factor to the changes of resonance parameters of interest. In the present work, the uncertainties of the resolved resonance parameters for the evaluated nuclear data file JENDL-3.2 were estimated on the basis of Breit-Wigner Multi-level formula. The resonance self-shielding factor based on NR-approximation is analytically described. Reactivity uncertainty evaluation method for the effective multiplication factor k
, temperature coefficient 
, and Doppler reactivity worth 
is developed by means of the sensitivity coefficient against the resonance parameter. Final uncertainties are estimated by means of error propagation law using the level-wise uncertainties. Preliminary uncertainty of Doppler reactivity worth results about 4% at the temperature 728 K for large sodium-cooled FBR.
Takada, Eiji*; Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Shimakawa, Satoshi; Nojiri, Naoki; Fujimoto, Nozomu
JAERI-Tech 2004-048, 60 Pages, 2004/06
JAERI-Tech-2004-048.pdf:4.18MB
The HTTR (High Temperature Engineering Test Reactor), which has thermal output of 30MW, coolant inlet temperature of 395
C and coolant outlet temperature of 850C/950C, is a first high temperature gas-cooled reactor (HTGR) in Japan. The HTGR has a high inherent safety potential to accident condition. Safety demonstration tests using the HTTR are underway in order to demonstrate such excellent inherent safety features of the HTGR. The reactivity insertion test demonstrates that rapid increase of reactor power by withdrawing the control rod is restrained by only the negative reactivity feedback effect without operating the reactor power control system, and the temperature transient of the reactor is slow. The best estimated analyses have been conducted to simulate reactor transients during the reactivity insertion test. A one-point core dynamics approximation with one fuel channel model is applied to this analysis. It was found that the analytical model for core dynamics could simulate the reactor power behavior.
Nakagawa, Shigeaki; Sakaba, Nariaki; Takada, Eiji*; Tachibana, Yukio; Saito, Kenji; Furusawa, Takayuki; Sawa, Kazuhiro
JAERI-Tech 2003-049, 22 Pages, 2003/03
JAERI-Tech-2003-049.pdf:1.17MB
Safety demonstration tests in the HTTR (High Temperature Engineering Test Reactor) will be carried out in order to verify inherent safety features of the HTGR (High Temperature Gas-cooled Reactor). The first phase of the safety demonstration tests includes the reactivity insertion test by the control rod withdrawal and the coolant flow reduction test by the gas circulator trip. In the second phase, accident simulation tests will be conducted. By comparison of their experimental and analytical results, the prediction capability of the safety evaluation codes such as the core and the plant dynamics codes will be improved and verified, which will contribute to establish the safety design and the safety evaluation technologies of the HTGRs. The results obtained through its safety demonstration tests will be also utilised for the establishment of the safety design guideline, the safety evaluation guideline, etc. This paper describes the test program of the overall safety demonstration tests and the test method, the test conditions and the results of the pre-test analysis of the reactivity insertion test and the partial gas circulator trip test planned in March 2003.
Nakajima, Ken; ;
Proceedings of 6th International Conference on Nuclear Criticality Safety (ICNC '99), 3, p.1286 - 1292, 1999/00
no abstracts in English
*; Ohashi, Kazutaka*; *; *; Sawa, Kazuhiro; Nakagawa, Shigeaki; Kunitomi, Kazuhiko
FAPIG, 0(127), p.11 - 19, 1991/03
no abstracts in English
for thermal fission of 
U based on integral experiments at using SHE; Akino, Fujiyoshi;
Journal of Nuclear Science and Technology, 21(6), p.487 - 488, 1984/00
Times Cited Count：3 Percentile：59.52(Nuclear Science & Technology)no abstracts in English
