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Sawaguchi, Takuma
no journal, ,
no abstracts in English
Takahara, Shogo
no journal, ,
Dose assessment is important for providing appropriate protection to the people and clarifying the impact of the accident. To assess the doses realistically and comprehensively, a probabilistic approach was adopted using data that reflected realistic environmental trends and lifestyle habits. In the first year after the contamination, the 95th percentile of the annual effective dose received by the inhabitants evacuated from the evacuation areas and the deliberate evacuation areas was mainly in the 1-10 mSv dose band. The 95th percentile of the dose received by some outdoor workers and inhabitants evacuated from highly contaminated areas was in the 10-50 mSv dose band. The doses due to external exposure to deposited radionuclides were the dominant exposure pathway, and their contributions were about 90%. In addition, 20-30% of the lifetime effective dose was delivered during the first year.
Takeda, Takeshi
no journal, ,
The OECD/NEA ROSA Project has been conducted in the thermal-hydraulic safety research group of JAEA for 7.5 years from April 2005 to resolve reactor safety issues. Nineteen LSTF tests simulated nine subjects on reactor accidents. Researches have been done to intend improvements of thermal-hydraulic best-estimate methods and identification of the phenomena based on obtained test data. Typical major results are introduced for separate-effect test on steam condensation in cold legs during large-break LOCA and system-effect tests on cold leg intermediate-break LOCAs. A series of blind analyses for the intermediate-break LOCA tests were performed with Project participants, which caused significant scattering in the peak cladding temperature and core liquid level. Analysis models such as CCFL at core exit, which need for prediction improvement of core cooling, were clarified through the post-test analyses.
Kimura, Masanori; Ishikawa, Jun
no journal, ,
no abstracts in English
Iida, Yoshihisa
no journal, ,
For the long term assessment, it is important to evaluate the variation of distribution coefficients depending on the various conditions of geological medium and groundwater composition. The method for setting sorption distribution coefficients and their variation ranges was developed for the safety assessment of geological disposal, by taking into account the mechanistic sorption model calculation and the distribution coefficient data acquired by using geologic samples collected from disposal site, in addition to the existing distribution coefficient data acquired under various conditions.
Yamato, Masaaki; Sugiyama, Tomoyuki; Nagase, Fumihisa
no journal, ,
To confirm the reactor safety design, safety reviews are performed under accident conditions as well as under normal operation conditions. One of such Design Basis Accidents is the Loss-of-Coolant Accident (LOCA) in which the reactor coolant water is lost typically due to a pipe break. This presentation shows the concept of the current LOCA safety regulation, and presents the results from the recently started experiments on mechanical properties of post-LOCA cladding, breakaway phenomenon in cladding high temperature oxidation, and influence of sea water on the cladding high temperature oxidation.
Tobita, Toru; Udagawa, Makoto; Katsuyama, Jinya; Onizawa, Kunio; Nishiyama, Yutaka
no journal, ,
For safe operation of light water reactors, the structural integrity of reactor pressure vessel (RPV) is confirmed based on the evaluation and prediction of aging degradation such as irradiation embrittlement and stress corrosion cracking. To contribute to this issues, we have performed to study on the improvement of fracture toughness evaluation method and development of probabilistic fracture mechanics (PFM) analysis code for RPVs. It was clarified that temperature dependence of the fracture toughness of RPV steels with different toughness levels can be evaluated with enough accuracy using miniature specimens. In addition, in order to evaluate the failure probability of the RPV head penetration due to the PWSCC at the nickel based dissimilar metal welds, we have developed a PFM analysis code PASCAL-NP. The major findings of these studies are summarized in the poster.
Katsuyama, Jinya; Yamaguchi, Yoshihito; Onizawa, Kunio; Nishiyama, Yutaka
no journal, ,
Japanese nuclear power plants experienced large earthquakes, such as the Tohoku District-off the Pacific Ocean Earthquake in 2011. The magnitude of the earthquake exceeded the design-base condition for some reactor components. In addition, the Japanese seismic design guideline revised in 2006 stipulates that the residual risk should be considered assuming that seismic load exceeds design-base condition. However, it is not decided how to evaluate the residual risk. Therefore, it is an important issue to establish the evaluation method of excessive loading on the structural integrity of reactor components. In this presentation, a crack growth evaluation method based on an elastic-plastic fracture mechanics is presented as one of our research results. Future plans of our researches to evaluate the effect of large earthquake on the structural integrity of reactor components based on the probabilistic fracture mechanics analysis and three-dimensional large scale analysis are also presented.
Satou, Akira
no journal, ,
The objective of the plan of thermohydraulic research considering the severe accident (SA) is the development and advancement for thermohydraulic safety evaluation methods.
Abe, Hitoshi; Tashiro, Shinsuke; Yamane, Yuichi; Amano, Yuki; Yoshida, Kazuo; Uchiyama, Gunzo
no journal, ,
To establish the probabilistic safety assessment (PSA) methodology for fuel reprocessing plant, acquisition of data for evaluating the phenomenon that the frequency is very low but the influence will be large is important subject. As the collaboration study among three organizations of JAEA, JNES and JNFL, the release and transport characteristics of radioactive materials at boiling accident of concentrated high level radioactive solution waste (CHLW) in reprocessing plants has been investigated. In the study three kind of tests are conducted such a small scale cold test, engineering scale cold test and small scale hot test. In the small scale cold test, the release and transport characteristics of simulated FP from simulated CHLW heated were examined. In this work, outline of this study and example results of small scale cold test will be presented.
