Research on consequence analysis method for probabilistic safety assessment of nuclear fuel facilities, 2; Basic experimental data aerosol release fraction and trial analysis in boiling event of radioactive solution

Yoshida, Kazuo; Hayashi, Kazuya*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 9(1), p.60 - 70, 2010/03

https://doi.org/10.3327/taesj.J09.003

A special committee of "Research on the analysis methods for accident consequence of nuclear fuel cycle facilities (NFCF)" was organized by the Atomic Energy Society of Japan under the entrustment of Japan Atomic Energy Agency for a research on the state-of-the-art of consequence analysis method for Probabilistic Safety Assessment (PSA) of NFCF such as fuel reprocessing and fuel fabrication facilities. The objectives of this research is to obtain the basic useful information related for establishing the quantitative performance requirement and for risk-informed regulation through qualifying issues needed to be resolved. The research activities were mainly focused on accidents with more severe consequence than design basis such as events of criticality, explosion, fire and boiling of radioactive solution. The research results are summarized in this technical report about basic experimental data related to key phenomena postulated in boiling event of radioactive solution.

Research on the state-of-the-art of probabilistic safety assessment for non-reactor nuclear facilities (II)

Yoshida, Kazuo; Abe, Hitoshi; Yamane, Yuichi; Tashiro, Shinsuke; Muramatsu, Ken

JAEA-Research 2007-002, 127 Pages, 2007/03

JAEA-Research-2007-002.pdf:6.64MB

Japan Atomic Energy Agency (JAEA) entrusted with a research on the state-of-the-art of probabilistic safety assessment (PSA) of non-reactor nuclear facilities (NRNF) such as fuel reprocessing and fuel fabrication facilities to the Atomic Energy Society of Japan (AESJ). The objectives of this research is to obtain the basic useful information related for establishing the quantitative performance requirement and for risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NRNF. A special committee of "Research on the analysis methods for accident consequence in NFRF" was organized by the AESJ. The research activities of the committee were mainly focused on the analysis method for upper bounding consequences of accidents such as events of criticality, explosion, fire and solvent boiling postulated in NRNF resulting in release of radio active material to the environment. This report summarizes the results of research conducted by the committee in FY 2005.

Report of special committee on reactor physics code system for joint use

*

JNC TJ9420 2001-007, 167 Pages, 2001/03

JNC-TJ9420-2001-007.pdf:6.07MB

Up to the present, reactor physics codes and data which have been developed in each university and public institutein our country have been managed by each institution individually, and there have been no circumstances that they could be used commonly and conveniently. On account of this situation, a considerable amount of work had to be spent for the improvement of input and output procedures of codes and the development of interfaces between codes. Calculations for the verification of new developments and improvements of codes have been also made individually by developers or users themselves, and such information has been seldom owned jointly. On the other hand, the efforts to build the environment for implementing and using codes which werc originally developed in universities and public institutes commonly, have been made for many years in Europe and US, and especially in Europe, a reactor physics code system has been developed in the framework of an international cooperation of Great Britain, France, Germany et al. beyond the boundaries of Nations. By developing a common reactor physics code system which can be used for a wide area of applications, and organizing standard formats and interfaces for many reactor physics codes developed in our country, the Japanese research environment will be improved significantly and our research activities and capabilities will be increased. Exchange of data libraries and of input and output data will be facilitated so that cooperation and sharing of work can be improved. Further, using information technology, information related to the current code system, distribution of code manuals, distributed development of the code system and checking of common data for verifications and validations may be done more efficiently. In the present committee, in order to investigate a possibility of development of such a reactor physics code system, getting together code developer and users, investigations and discussions were made ...

Methodology for assessment on safety performance of geological disposal of high-level wastes

Atomic Energy Society of Japan*

JNC TJ1400 2005-011, 199 Pages, 1988/07

JNC-TJ1400-2005-011.pdf:6.88MB

One of the most crucial elements for assessing the safety of geological disposal of high-level radioactive waste is how to deal with uncertainties associated with its long-term and long-distance natures.

State-of-the-art and key issues of accident consequence analysis method for nuclear fuel facilities

Yoshida, Kazuo

no journal, , 

no abstracts in English

Basic principles of maintenance management for R&D-level nuclear power plants

Yamaguchi, Akira*; Ito, Takaya*; Takaya, Shigeru; Kotake, Shoji*

no journal, , 

no abstracts in English