Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Ishikawa, Jun; Muramatsu, Ken; Sakamoto, Toru*
JAERI-Research 2005-021, 133 Pages, 2005/09
JAERI-Research-2005-021.pdf:7.58MB
The THALES-2 code is an integrated severe accident analysis code in order to simulate the accident progression and transport of radioactive material for probabilistic safety assessment (PSA) of a nuclear power plant, a part of a level 3 PSA being performed at JAERI for a 1,100MWe BWR-5 with a Mark-II containment. Results and insights from the analyses were that (1) the calculated release fractions of CsI and CsOH to the environment were in the range of 0.01 to 0.1 for late containment overpressure failure cases, and the release fractions for the containment venting case were one order of magnitude smaller than that of over-pressure case and those for drywell spray recovery cases where no containment failure occurred were two orders of magnitude smaller than the containment venting cases, (2) the governing factors for source terms of Iodine and Cesium are different depending on whether the containment fails before core melt or not, (3) the containment venting, which is one of the accident management measures, can be expected to reduce source terms if suppression pool bypass is avoided.
Moriyama, Kiyofumi; Takagi, Seiji; Muramatsu, Ken; Nakamura, Hideo; Maruyama, Yu
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 9 Pages, 2005/05
The containment failure probability due to ex-vessel steam explosions were evaluated for a BWR Mk-II model plant. The evaluation was made for two scenarios: a steam explosion in the pedestal area, or in the suppression pool. A probabilistic approach, Latin Hypercube Sampling (LHS), was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The fragility curves connecting the steam explosion loads and containment failure were developed based on simplified assumptions on the containment failure scenarios. The mean conditional probabilities of containment failure per occurrence of a steam explosion were 
for suppression pool and 
for pedestal area. Note that the results depend on the assumed range of input parameters and fragility curves that involve conservatism and simplification.
21, 1997, Tokai-mura, Japan; ; Sugimoto, Jun
JAERI-Conf 97-011, 829 Pages, 1998/01
no abstracts in English
Kudo, Tamotsu; Yamano, N.; Moriyama, Kiyofumi; Maruyama, Yu; Sugimoto, Jun
3rd Int. Conf. on Containment Design and Operation,Conf. Proc., Vol. 1, 0, 10 Pages, 1994/00
no abstracts in English