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Onoda, Yuichi; Kurisaka, Kenichi; Sakai, Takaaki
Journal of Nuclear Science and Technology, 53(11), p.1774 - 1786, 2016/11
Times Cited Count：3 Percentile：28.28(Nuclear Science & Technology)Allelein, H.-J.*; Auvinen, A.*; Ball, J.*; G
ntay, S.*; Herranz, L. E.*; Hidaka, Akihide; Jones, A. V.*; Kissane, M.*; Powers, D.*; Weber, G.*
NEA/CSNI/R(2009)5, 388 Pages, 2009/12
Watanabe, Yuichi*; Muramatsu, Ken; Oikawa, Tetsukuni
Nuclear Engineering and Design, 235(23), p.2495 - 2512, 2005/12
Times Cited Count：2 Percentile：17.6(Nuclear Science & Technology)This paper presents an evaluation of seismic capacity of a group of vertical U-tube type heat exchangers(HXs) with support frames for residual heat removal systems of BWRs for seismic Probabilistic Safety Assessment in Japan. The median capacity was evaluated by a time history response analysis with a detailed model for a representative HX selected from four HXs. The logarithmic standard deviation(LSD) for uncertainty due to lack of knowledge was evaluated with consideration of the variabilities in three influential parameters, i.e., diameter of anchor bolt, weight of HX and position of center of gravity of HX. The dominant failure mode of HXs was the failure of anchor bolts of lugs mainly due to shearing stress. The capacity expressed in terms of zero period acceleration on the foundation of HX was evaluated to be 4,180 Gal(4.3 g) for median, LSD for uncertainty due to randomness was 0.11 from the variability in material property and LSD due to lack of knowledge was 0.21 to 0.53 depending on combination of the variability in design parameters to be considered.
Tamaki, Hitoshi; Hamaguchi, Yoshikane; Yoshida, Kazuo; Muramatsu, Ken
Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10
A PSA procedure for MOX fuel fabrication facilities is being developed at the JAERI. This procedure consists of four steps, which are hazard analysis, accident scenario analysis, frequency evaluation and consequence evaluation. The proposed procedure is characterized by the hazard analysis step. The Hazard analysis step consists of two sub-steps. In the first sub-step, a variety of functions of equipment composing the facility are analyzed to identify potential abnormal events exhaustively. In the second sub-step, these potential events are screened to select abnormal events by using a risk matrix based on the rough estimation of likelihood and maximum unmitigated release of radioactive material. One of the unique technical issues in this research is the estimation of likelihood of criticality event. A method is also proposed as a part of PSA procedure taking into consideration of failure of a computerized control system for MOX powder handling process. The applicability of the PSA procedure was demonstrated through the trial application of it to a model plant of MOX fuel fabrication facility.
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.
Homma, Toshimitsu; Tomita, Kenichi*; Hato, Shinji*
Nuclear Engineering and Technology, 37(3), p.245 - 258, 2005/06
This paper addresses two types of uncertainty: stochastic uncertainty and subjective uncertainty in probabilistic accident consequence assessments. The off-site consequence assessment code OSCAAR has been applied to uncertainty and sensitivity analyses on the individual risks of early fatality and latent cancer fatality in the population due to a severe accident. A new stratified meteorological sampling scheme was successfully implemented into the trajectory model for atmospheric dispersion and the statistical variability of the probability distributions of the consequence was examined. A total of 65 uncertain input parameters was considered and 128 runs of OSCAAR were performed in the parameter uncertainty analysis. The study provided the range of uncertainty for the expected values of individual risks of early and latent cancer fatality close to the site. In the sensitivity analyses, the correlation/regression measures were useful for identifying those input parameters whose uncertainty makes an important contribution to the overall uncertainty for the consequence.
Hidaka, Akihide; Kudo, Tamotsu; Ishikawa, Jun; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 42(5), p.451 - 461, 2005/05
Times Cited Count：6 Percentile：40.47(Nuclear Science & Technology)The radionuclide release from MOX under severe accident conditions was investigated in VEGA program to contribute to the technical bases for safety evaluation including PSA for LWR using MOX. The MOX specimens irradiated at ATR Fugen were heated up to 3123K in helium at 0.1 and 1.0MPa. The release of volatile FP was slightly enhanced below 2200K compared with that of UO
. The volatile FP release at elevated pressure was decreased as in the case with UO. The total fractional release of Cs reached almost 100% while almost no release of low-volatile FP even after the fuel melting. The release rate of plutonium above 2800K increased rapidly although the amount was small. Since the existing models cannot predict this increase, an empirical model was prepared based on the data. There is no large difference in FP inventories between UO and MOX, and the fractional releases from MOX can be mostly predicted by the model for UO. This suggests that the consequences of LWR using MOX are mostly equal to those using UO from a view point of risks.
Watanabe, Norio
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(4), p.396 - 406, 2004/12
The U. S. Nuclear Regulatory Commission has been carrying out the Accident Sequence Precursor (ASP) Program to identify and categorize precursors to potential severe core damage accident sequences using the PSA technique. The ASP Program has identified a lot of risk significant events as precursors that occurred at U. S. nuclear power plants. The results from the ASP Program include valuable information that could be useful for obtaining risk significant insights and for monitoring risk trends in nuclear power industry. The present study examines and discusses quantitative risk trends for the industry level, using two indicators, that is, the occurrence frequency of precursors and the annual core damage probability, deriving from the ASP analysis results. It is shown that the core damage risk at U.S. nuclear power plants has been lowered and the likelihood of risk significant events has been remarkably decreasing. Also, the present study demonstrates that two risk indicators used here can provide quantitative information useful for monitoring risk trends in nuclear power industry.
Shimizu, Akira; Nishihara, Tetsuo; Moriyama, Koichi*
JAERI-Tech 2004-051, 69 Pages, 2004/06
JAERI-Tech-2004-051.pdf:4.68MB
HTTR of JAERI will be connected with a hydrogen production system by steam reforming of methane for development of nuclear heat utilization technology. This facility will handle much inflammable gas near the nuclear reactor so that special safety consideration is necessary. This report describes the Probabilistic Safety Assessment (PSA) of inflammable gas leakage in the HTTR hydrogen production system. Vessels and pipes, which contain flammable gas, were divided into several systems. Probability of gas leakage were calculated at all candidate places. As a result of assessment, the counter measures such as double-covered inflammable gas pipes, small diameter instrument pipes, leakage detector and emergency shut off valves, are confirmed to be very effective to minimize the scale of explosion and to prevent the damage on nuclear plant.
Tsutsumi, Hideaki*; Ebisawa, Katsumi*; Yamada, Hiroyuki*; Shibata, Katsuyuki; Fujimoto, Shigeru*
Nihon Zairyo Gakkai JCOSSAR 2003 Rombunshu, p.829 - 836, 2003/11
no abstracts in English
Homma, Toshimitsu
Proceedings of 4th International MACCS Users Group Meeting, p.57 - 66, 2002/10
no abstracts in English
Takeda, Seiji; Kimura, Hideo
JAERI-Research 2002-014, 20 Pages, 2002/07
JAERI-Research-2002-014.pdf:1.37MB
no abstracts in English
Hirose, Jiro*; Muramatsu, Ken; Okumura, Toshihiko*; Taki, Satoshi*
JAERI-Research 2002-009, 220 Pages, 2002/03
JAERI-Research-2002-009.pdf:13.31MB
This report presents procedures for the determination of Scenario Earthquakes for seismic design based on Probabilistic Seismic Hazard Analysis (PSHA). Recently PSHA was recognized as an important basis to identify dominant earthquakes predicted to threaten the site in future. The identified earthquakes are called Probability-Based Scenario Earthquakes (PBSEs). The concept of PBSEs originates from the studies of US NRC and Ishikawa & Kameda. The objective of this study is to formulate the procedures to determine the PBSEs and, through this application, to demonstrate the feasibility of the application to seismic design. This report consists of three parts, namely, procedures to compile analytical conditions for PBSEs, an assessment to determine PBSEs for a model site using the Ishikawa's concept and examination of uncertainty involved in analytical conditions. The results imply that the procedures based on the Ishikawa's concept is a useful evaluation technique to determine scenario earthquakes for seismic design considering uncertainty involved in analytical conditions.
Liu, X.*; Tomita, Kenichi*; Homma, Toshimitsu
JAERI-Research 2002-004, 37 Pages, 2002/03
JAERI-Research-2002-004.pdf:2.14MB
One important step in Level 3 Probabilistic Safety Assessment is meteorological sequence sampling, on which the previous studies were mainly related to code systems using straight line plume model and more efforts are needed for trajectory puff model such as the OSCAAR code system. This report describes the development of a new meteorological sampling scheme for the OSCAAR code system that explicitly considers population distribution. A group of principles was set forth for the development of this new sampling scheme, including completeness, stratification, sample allocation, practicability and so on. The calculation results illustrate that although it is quite difficult to idealize stratification of meteorological sequences based on a few environmental parameters the new scheme do gather the most inverse conditions in a single subset of meteorological sequences. The size of this subset may be as small as a few dozens, so that the tail of a CCDF curve is possible to remain relatively static in different trials of the PCA code system.
Yamada, Hiroyuki; Tsutsumi, Hideaki*; Ebisawa, Katsumi*; Suzuki, Masahide
JAERI-Data/Code 2002-001, 161 Pages, 2002/03
JAERI-Data-Code-2002-001.pdf:6.62MB
no abstracts in English
Yokobayashi, Masao; Oikawa, Tetsukuni; Muramatsu, Ken
Nihon Genshiryoku Gakkai Wabun Rombunshi, 1(1), p.95 - 105, 2002/03
no abstracts in English
Ishikawa, Jun; Shintani, Kiyonori; Takagi, Seiji; Muramatsu, Ken
Nihon Kikai Gakkai Dai-8-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.203 - 208, 2002/00
no abstracts in English
Ishikawa, Jun; Muramatsu, Ken; Sakamoto, Toru*
Proceedings of 10th International Conference on Nuclear Engineering (ICONE 10) (CD-ROM), 8 Pages, 2002/00
no abstracts in English
Tamaki, Hitoshi; Watanabe, Norio*; Muramatsu, Ken
Proceedings of the 2001 Topical Meeting on Practical Implementation of Nuclear Criticality Safety (CD-ROM), 10 Pages, 2001/11
no abstracts in English
Liu, X.; Homma, Toshimitsu
JAERI-Tech 2001-054, 49 Pages, 2001/08
JAERI-Tech-2001-054.pdf:1.54MB
no abstracts in English
