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Futagami, Satoshi; Yamano, Hidemasa; Kurisaka, Kenichi; Ujita, Hiroshi*
Proceedings of proceedings of PSAM 2023 Topical Conference AI & Risk Analysis for Probabilistic Safety/Security Assessment & Management, 8 Pages, 2023/10
To create an innovation for efficient and effective social implementation of nuclear power plant PRA, automatic construction tool for fault tree architecture and automatic failure judgment tool to construct reliability database are developed by using AI and digitization technology. This paper describes overall development plan of PRA methodology using the AI technology and the progress of automatic FT creation tools development.
Shimada, Taro; Namekawa, Masakazu*; Miwa, Kazuji; Takeda, Seiji
Proceedings of Waste Management Symposia 2023 (WM2023) (Internet), 8 Pages, 2023/02
It is supposed that radioactive dust deposited at the land surface will be moved downstream and concentrated at the depression by overland flow at heavy rain after the accidental release of radioactive dusts accumulated at the filters in the decommissioning stage of nuclear facilities. The authors are developing a calculation code to evaluate distribution changes of radioactivity on the surface and public dose considering the conditions such as rainfall, topography and types of cover surface. It is necessary to construct methods for setting parameter values used for the calculations based on the actual situation. Therefore, the parameter values were obtained by the experiments where Fe
O
powder spread on the cover surface such as smooth and aged-asphalt, concrete and bare soil, was eroded by overland flow and raindrops and they were collected at the lower end of the slope at a minute interval. The collected weights of overland flow and FeO powder were measured. Based on the Manning's roughness coefficient for smooth asphalt already known as a fixed value, the erosion velocity coefficient was evaluated. Then Manning's roughness coefficients for other cover surfaces were obtained using the erosion velocity coefficient. Manning's roughness coefficients were slightly smaller than the range of literature values. In addition, elevations for the cover surface were measured by 3D scanner as point cloud data, and the surface roughness were evaluated. The values of surface roughness and Manning's roughness coefficients had a correlation. It indicated a possibility to utilize the surface roughness to set the Manning's roughness coefficient for the evaluation of radioactivity distribution change by heavy rainfall.
Kinoshita, Takahiro*; Okamura, Shigeki*; Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi; Futagami, Satoshi; Fukasawa, Tsuyoshi*
Transactions of the 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 7 Pages, 2022/07
The seismic evaluation of key components such as reactor vessel is important for the Seismic Probabilistic Risk Assessment (S-PRA) in a Sodium-Cooled Fast Reactor (SFR). Many components were damaged by cumulative damage like fatigue damage during seismic ground motion. However, general evaluation method for key components under seismic ground motion has been based on static loads and elastic region of materials. More accurate evaluation method for S-PRA, which can evaluate the failure of key components such as reactor vessels, has been actually required. In this study, failure probability evaluation method with integrated energy was developed by comparing the energy with vibration tests and fatigue tests. Vibration tests were performed to evaluate integrated vibration energy at failure by energy balance equation and fatigue tests were performed to evaluate integrated vibration energy at failure based on experimental results of fatigue tests.
Yamaguchi, Yoshihito; Li, Y.
Haikan Gijutsu, 63(12), p.22 - 27, 2021/10
no abstracts in English
Maruyama, Yu; Kita, Toshinobu*; Kuramoto, Takahiro*
Nihon Genshiryoku Gakkai-Shi ATOMO
, 62(6), p.328 - 333, 2020/06
no abstracts in English
Sato, Hiroyuki; Nishida, Akemi; Ohashi, Hirofumi; Muramatsu, Ken*; Muta, Hitoshi*; Itoi, Tatsuya*; Takada, Tsuyoshi*; Hida, Takenori*; Tanabe, Masayuki*; Yamamoto, Tsuyoshi*; et al.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
JAEA, in conjunction with Tokyo City University, The University of Tokyo and JGC Corporation, have started development of a PRA method considering the safety and design features of HTGR. The primary objective of the project is to develop a seismic PRA method which enables to provide a reasonably complete identification of accident scenario including a loss of safety function in passive system, structure and components. In addition, we aim to develop a basis for guidance to implement the PRA. This paper provides the overview of the activities including development of a system analysis method for multiple failures, a component failure data using the operation and maintenance experience in the HTTR, seismic fragility evaluation method, and mechanistic source term evaluation method considering failures in core graphite components and reactor building.
Matsuda, Kosuke*; Muramatsu, Ken*; Muta, Hitoshi*; Sato, Hiroyuki; Nishida, Akemi; Ohashi, Hirofumi; Itoi, Tatsuya*; Takada, Tsuyoshi*; Hida, Takenori*; Tanabe, Masayuki*; et al.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
This paper proposes a set of procedures for accident sequence analysis in seismic PRAs of HTGRs that can consider the unique accident progression characteristics of HTGRs. Main features of our proposed procedure are as follows: (1) Systematic analysis techniques including Master Logic Diagrams are used to ensure reasonable completeness in identification of initiating events and classification of accident sequences, (2) Information on factors that govern the accident progression and source terms are effectively reflected to the construction of event trees for delineation of accident sequences, and (3) Frequency quantification of seismically-initiated accident sequence frequencies that involve multiplepipe ruptures are made with the use of the Direct Quantification of Fault Trees by Monte Carlo (DQFM) method by a computer code SECOM-DQFM.
Takata, Takashi; Azuma, Emiko*
Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 10 Pages, 2016/10
A new approach has been developed to assess event sequences under external hazard considering a plant status quantitatively and stochastically so as to take various scenarios into account automatically by applying a Continuous Markov Chain Monte Carlo (CMMC) method coupled with a plant dynamics analysis. In the paper, a tornado and a strong wind are selected as the external hazard to assess the plant safety in a loop type sodium cooled fast reactor (SFR). As a result, it is demonstrated that the various scenarios where the order of the occurrence event and its occurrence time differs from each other can be assessed simultaneously as well as the statistical characteristics of plant parameter such as the coolant temperature. Furthermore, a weight factor is introduced so as to investigate the low failure probability events with a comparative small number of the sampling.
Yamano, Hidemasa; Nishino, Hiroyuki; Okano, Yasushi; Yamamoto, Takahiro*; Takata, Takashi*
Earthquakes, Tsunamis and Nuclear Risks, p.111 - 121, 2016/01
The present study is developing risk assessment methodologies that include probabilistic risk assessment (PRA) and margin assessment methodologies against snow, tornado, strong wind, rain, volcanic eruption and forest fire mainly for a sodium-cooled fast reactor. The present paper describes briefly the project overview and then mainly the development of PRA and margin assessment methodologies against strong wind. In the strong wind PRA, the hazard curve was estimated using the Gumbel distributions based on weather data. Next, failure probabilities for components were calculated and event trees were developed. Using them, the strong wind PRA methodology was developed to quantify a core damage frequency. The present study also developed the wind margin assessment methodology that the margin was regarded as wind speed leading to the decay heat removal failure.
Okano, Yasushi; Yamano, Hidemasa
Proceedings of International Topical Meeting on Probabilistic Safety Assessment and Analysis (PSA 2015) (USB Flash Drive), p.22 - 31, 2015/04
An external hazard curve of a forest fire is evaluated based on a logic tree. The logic tree consists domains of "forest fire breakout and spread conditions", "weather condition" and "vegetation and topographical conditions". A location nearby a typical nuclear power plant site in Japan was selected, and the frequency of a large forest fire of the location is approximately 1/5 of the average in Japan. Forest fire breakout points were selected considering typical forest fire causes in Japan. The weather conditions are represented by two parameter sets of "temperature-humidity" and "wind direction-wind speed". A number of forest fire simulations were performed to obtain a response surface for a frontal fireline intensity. The hazard curve is therefore evaluated by a Monte Carlo simulation such that the annual exceedance probability is about 1.0
10
per year for the frontal fireline intensity of 200 kW/m and about 1.310
per year for 300 kW/m.
Takahara, Shogo; Hidaka, Akihide; Ogino, Takashi*
JAEA-Data/Code 2015-001, 65 Pages, 2015/03
JAEA-Data-Code-2015-001.pdf:10.74MB
HEINPUT code is one of the preprocessor for probabilistic accident consequence assessment code OSCAAR, and estimates the risk of incidence and death due to radiation induced cancers. HEINPUT code currently uses two models developed by U.S. Nuclear Regulatory Commission (USNRC, 1985; 1993) and U.S. Environmental Protection Agency (EPA, 1994). In this report, the code was improved to enable to make calculation using the new EPA model. In addition, in order to reduce user's burden, we developed input data generator which can provide the input data for three estimation models implemented in HEINPUT-GUI based on the statistical information published.
transformation; Article reviewSaigusa, Shin*
JAERI-Review 2002-009, 43 Pages, 2002/03
JAERI-Review-2002-009.pdf:2.3MB
The mechanism of the process of radiation-induced transformation that leads normal mammalian cell to malignant feature is still not clear. However, a drastic development of molecular biological technique in medical biology research field since 1990's have progressively revealed the mechanisms of this process and recent results have been starting to incorporate in the recent radiation risk estimate procedures. This report describes the results of the survey and review of the articles concerned with experimental studies of radiation induced in vitro transformation, since 1993 to 1997.
Saigusa, Shin*
RIST News, (32), p.2 - 11, 2001/10
The Japan Atomic Energy Research Institute (JAERI) had carried out the specially funded project of constructing a database "DRESA (Database for Radiation Exposure and Safety Assessment)" with an electronic text on safety assessment of low dose ionizing radiation. The contents and the database system were arranged and designed to provide the fundamental information/data for scientific and reasonable understanding of radiation health effects, radiation risk assessment and radiation protection. Present report explained the data title and its contents, data file formats and system structure of the database. Potentialities and difficulties in constructing WWW database were also discussed.
Mizushita, Seiichi
Dai-39-kai Genshiryoku Sogo Shimpojium Yokoshu, p.33 - 41, 2001/04
no abstracts in English
Hidaka, Akihide
RIST News, (30), p.2 - 14, 2000/10
no abstracts in English
Tamura, Kazuo*; Iriya, Yoshikazu*
JNC TJ9440 2000-004, 22 Pages, 2000/03
JNC-TJ9440-2000-004.pdf:2.35MB
In the probabilistic safety assessment(PSA), the fault tree/event tree technique has been widely used to evaluate accident sequence frequencies. However, event tansition which operators actually face can not be dynamically treated by the conventional technique. Therefore, we have made the dynamic analysis program(DYANA) for event transition for a liquid metal cooled fast breeder reactor. In the previous development, we made basic model for analysis. However, we have a probrem that calculation time is too long. At the current term, we made parallelization of DYANA usig MPI. So we got good performance on WS claster. It performance is close to ideal one.
*; *
JNC TJ9440 2000-002, 90 Pages, 2000/03
JNC-TJ9440-2000-002.pdf:1.43MB
In order to support development of the dynamic reliability analysis program DYANA, analyses were made on the event sequences anticipated under emergency situations using the plant dynamics simulation computer code Super-COPD. In this work 9 sequences were analyzed and integrated into an input file for preparing the functions for DYANA using the analytical model and input data which developed for Super-COPD in the previous work. These sequences could not analyze in the previous work, which were categorized into the PLOHS (Protected Loss of Heat Sink) event.
Ikeda, Takao*; Yoshida, Hideji*; Miki, Takahito*
JNC TJ8400 2000-046, 264 Pages, 2000/02
JNC-TJ8400-2000-046.pdf:6.73MB
This report contains discussions about methodology for the selection of parameter values, stochastic approach for the biosphere assessment and biosphere modelling for marine discharge case are described. Regarding the methodology for the selection of parameter values, important aspects for the data selection were discussed, and data selection protocol was developed. Regarding the stochastic approach for the biosphere assessment, it is confirmed that Straightforward Monte Carlo Method and Latin Hypercube Sampling Method are the most adequate based on a literature survey. Then stochastic assessment by using biosphere model that was developed in the second progress report was carried out to check the sensitivity of parameter values. Finally, availability of several kind of assessment models for marine discharge case were discussed. It was confirmed that Multiple Compartment Model was the most applicable. Assessment using Multiple Compartment Model was carried out. The results were compared with those derived by numerical model. As a result, the difference between two models were small enough.
