Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 20
  • 1

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Level 1 PRA for external vessel storage tank of Japan sodium-cooled fast reactor in whole core refueling

Yamano, Hidemasa; Kurisaka, Kenichi; Nishino, Hiroyuki; Okano, Yasushi; Naruto, Kenichi*

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 15 Pages, 2018/10

Spent fuels are transferred from a reactor core to a spent fuel pool through an external vessel storage tank (EVST) filled with sodium in sodium-cooled fast reactors in Japan. This paper describes identification of dominant accident sequences leading to fuel failure, which was achieved through probabilistic risk assessment for the EVST designed for a next sodium-cooled fast reactor plant system in Japan to improve the EVST design. The safety strategy for the EVST involves whole core refueling (early transfer of all core fuel assemblies into the EVST) assuming a severe situation that results in sodium level reduction leading finally to the top of the reactor core fuel assemblies in a long time. This study introduces the success criteria mitigation along the decay heat decrease over time. Based on the design information, this study has carried out identification of initiating events, event and fault tree analyses, a probability analysis for human error, and quantification of accident sequences. The fuel damage frequency of the EVST was evaluated to be approx. 10$$^{-5}$$/year. The dominant accident sequence resulted from the static failure and human error for the switching from the stand-by to operation mode in the three stand-by cooling circuits after loss of one circuit for refueling heat removal operation as an initiating phase.

Journal Articles

Level 1 PRA for external vessel storage tank of Japan sodium-cooled fast reactor in scheduled refueling

Yamano, Hidemasa; Naruto, Kenichi*; Kurisaka, Kenichi; Nishino, Hiroyuki; Okano, Yasushi

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07

Spent fuels are transferred from a reactor core to a spent fuel pool through an external vessel storage tank (EVST) filled with sodium in sodium-cooled fast reactors in Japan. This paper describes identification of dominant accident sequences leading to fuel failure by conducting probabilistic risk assessment for EVST designed for a next sodium-cooled fast reactor plant system in Japan to improve the EVST design. Based on the design information, this study has carried out identification of initiating events, event and fault tree analyses, human error probability analysis, and quantification of accident sequences. Fuel damage frequency of the EVST was evaluated approx. 10$$^{-6}$$ /year in this paper. By considering the secondary sodium freezing, the fuel damage frequency was twice increased. The dominant accident sequence resulted from the common cause failure of the damper opening and/or the human error for the switching from the stand-by to the operation mode in the three stand-by cooling circuits. The importance analyses have indicated high risk contributions.

Journal Articles

Updating of local blockage frequency in the reactor core of SFR and PRA on consequent severe accident in Monju

Nishimura, Masahiro; Fukano, Yoshitaka; Kurisaka, Kenichi; Naruto, Kenichi*

Journal of Nuclear Science and Technology, 54(11), p.1178 - 1189, 2017/11

 Times Cited Count:2 Percentile:62.26(Nuclear Science & Technology)

Fuel subassemblies (FSAs) of fast breeder reactors (FBRs) are densely arranged and have high power densities. Therefore, PRA on LF which was initiated from LB was performed reflecting the state-of-the-art knowledge in this study. As the result, damage propagation from LF caused by LB in Monju can be negligible compared with the core damage due to ATWS or PLOHS in the viewpoint of both frequency and consequence.

Journal Articles

Level 1 PRA for external vessel storage tank of Japan sodium-cooled fast reactor in scheduled refueling

Yamano, Hidemasa; Naruto, Kenichi*; Kurisaka, Kenichi; Nishino, Hiroyuki; Okano, Yasushi

Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 3 Pages, 2017/11

Spent fuels are transferred from a reactor core to a spent fuel pool through an external vessel storage tank (EVST) filled with sodium in sodium-cooled fast reactors in Japan (JSFR). The objective of this study is to identify dominant accident sequences leading to fuel failure by conducting PRA for EVST. The EVST heat removal system in JSFR consists of four independent loops with for primary and secondary ones. Based on the JSFR design information, this study has identified initiating events, event and /fault tree analyses, human reliability analysis, and quantification of accident sequences. Fuel damage frequency of the EVST was evaluated approx. 10$$^{-6}$$ /year in this paper. The main contributor of the fuel damage frequency is the loss of heat removal function of the cooling system. The dominant initiating event was the loss of one circuit of normal heat removal operation.

Journal Articles

PRA on mixed foreign substances into core of Japanese prototype FBR

Nishimura, Masahiro; Fukano, Yoshitaka; Kurisaka, Kenichi; Naruto, Kenichi*

Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 12 Pages, 2016/10

Fuel subassemblies of fast breeder reactors (FBRs) are densely arranged and have high power densities. Therefore, the local fault (LF) has been considered as one of the possible initiating events of severe accidents. According to the LF evaluation under the condition of total flow blockage of one sub-channel in the analyses of design basis accident (DBA) for Monju, it was confirmed that the pin failures were limited locally without severe core damage. In addition, local flow blockage (LB) of 66% central planar in the subassembly was investigated as one of the beyond-DBA. However, it became clear that these deterministic analyses were not based on a realistic assumption by experimental studies. Therefore, PRA on LF which was initiated from LB was performed reflecting the state-of-the-art knowledge in this study. As the result, damage propagation from LF caused by LB in Monju can be included in CDF of ATWS or PLOHS in the viewpoint of both probability and consequence.

Journal Articles

Updating of adventitious fuel pin failure frequency in sodium-cooled fast reactors and probabilistic risk assessment on consequent severe accident in Monju

Fukano, Yoshitaka; Naruto, Kenichi*; Kurisaka, Kenichi; Nishimura, Masahiro

Journal of Nuclear Science and Technology, 52(9), p.1122 - 1132, 2015/09

 Times Cited Count:3 Percentile:62.51(Nuclear Science & Technology)

Experimental studies, deterministic approaches, and probabilistic risk assessments (PRAs) on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious-fuel-pin-failures (AFPFs) have been considered to be the most dominant initiators of LFs in these PRAs because of their high frequency of occurrence during reactor operation and possibility of fuel-element-failure-propagation (FEFP). A PRA on FEFP from AFPF (FEFPA) in the Japanese prototype SFR (Monju) was performed in this study based on the state-of-the-art knowledge, reflecting the most recent operation procedures under off-normal conditions. Frequency of occurrence of AFPF in SFRs which was the initiating event of the event tree in this PRA was updated using a variety of methods based on the above-mentioned latest review on experiences of this phenomenon. As a result, the frequency of occurrence of, and the core damage frequency (CDF) from AFPF in Monju was significantly reduced to a negligible magnitude compared with those in the existing PRAs. It was therefore concluded that the CDF of FEFPA in Monju could be comprised in that of anticipated-transient-without-scram or protected-loss-of-heat-sink events from both the viewpoint of occurrence probability and consequences.

Journal Articles

Seismic PRA for Japan Sodium-cooled Fast Reactor (JSFR)

Naruto, Kenichi*; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa; Okano, Yasushi; Okamura, Shigeki*; Eto, Masao*

Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 10 Pages, 2014/11

Journal Articles

Probability of adventitious fuel pin failures in fast breeder reactors and event tree analysis on damage propagation up to severe accident in Monju

Fukano, Yoshitaka; Naruto, Kenichi*; Kurisaka, Kenichi; Nishimura, Masahiro

Proceedings of 12th Probabilistic Safety Assessment and Management Conference (PSAM-12) (USB Flash Drive), 12 Pages, 2014/06

Experimental studies, deterministic and probabilistic and risk assessments (PRAs) on local fault (LF) propagation in sodium cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious fuel pin failures were considered to be the most dominant initiators of LFs in these PRAs because of high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. Therefore event tree analysis (ETA) on fuel element failure propagation initiated from adventitious fuel pin failure (FEFPA) in Monju was performed in this study based on state-of-the-art knowledge on experimental and analytical studies on FEFPA and reflecting latest operation procedure at emergency in Monju. Probability of adventitious fuel pin failures in SFRs which is the initiating event of this ETA was also updated in this study. It was clarified that FEFPA in Monju was negligible and could be included in core damage fraction of the anticipated transient without scram and protected loss of heat sink in the viewpoint of both probability and consequence.

Journal Articles

Estimation of component failure rates for PSA in sodium-cooled fast reactor

Naruto, Kenichi*; Kurisaka, Kenichi

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 9 Pages, 2012/12

Oral presentation

Failure probability of control rod insertion upon reactor scram of fast reactors

Kurisaka, Kenichi; Naruto, Kenichi*; Sugino, Tetsu*

no journal, , 

The failure probability of control rod insertion upon reactor scram is sensitive to the occurrence frequency of the reactor scram failure in sodium-cooled fast reactors (SFRs). The careful estimation of this probability is needed. The present study investigated the field data related to the scram demands and to their failure number. Data sources are "Joyo", "Monju", control rod drive mechanism mockup test that was performed in Japan, and PWRs in Japan and in US. Based on the investigation result, the failure probability of control rod insertion upon reactor scram was estimated for "Joyo" and "Monju" by using the hierarchical Bayes method with the Markov Chain Monte Carlo simulation. The failure probability was also estimated for a dummy reactor assuming a single success instance in rod insertion. The idea for the dummy reactor could be applied into the probability estimation of a new commercialized SFR having no operating experience.

Oral presentation

Development of component reliability database for LMFBRs

Naruto, Kenichi*; Kurisaka, Kenichi

no journal, , 

This study aims to implement a probabilistic safety assessment (PSA) based on the component operating experience in sodium-cooled fast reactors (SFR). For this purpose, we developed the component reliability database for LMFBRs named CORDS and evaluated the failure rate for SFR PSA using CORDS. The reliability data of 33 years and 15 years have been accumulated in CORDS from Joyo and Monju, respectively. Baysian method was applied to the failure rate evaluation. This evaluation was classified into three cases by considering applicable data amount: (1) when sufficient data of the target plant is available, conventional Bayesian update was applied by using its own data and non-informative prior; (2) when sufficient data of similar plants is available, parametric empirical Baysian method was applied; (3) when sufficient data is not available, conventional Bayesian update was implemented by using all data. This study served to evaluate the failure rate for Monju PSA.

Oral presentation

Development of component reliability database for LMFBRs, 2; Consistency of component boundary

Naruto, Kenichi*; Kurisaka, Kenichi

no journal, , 

In order to implement a probabilistic safety assessment (PSA) based on the component operating experience in sodium-cooled fast reactors (SFRs), we developed the component reliability database for SFRs named CORDS. For the purpose of consistency of the component boundary between CORDS and PSA analysis models, we extended the range of event data to collect further data of some components and evaluated the effect on the failure rate.

Oral presentation

Seismic PRA for Japan Sodium-cooled Fast Reactor (JSFR)

Naruto, Kenichi*; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa; Okano, Yasushi; Okamura, Shigeki*; Eto, Masao*

no journal, , 

no abstracts in English

Oral presentation

Level 1 PRA for spent fuel pool in Japan Sodium-cooled Fast Reactor (JSFR)

Naruto, Kenichi*; Sugino, Tetsu*; Yamano, Hidemasa; Kurisaka, Kenichi; Nishino, Hiroyuki; Okano, Yasushi

no journal, , 

no abstracts in English

Oral presentation

Evaluation of loss-of-offsite-power frequency and offsite power restoration

Miyabe, Takaaki*; Naruto, Kenichi*; Sugino, Tetsu*; Yamano, Hidemasa; Kurisaka, Kenichi; Nishino, Hiroyuki; Okano, Yasushi

no journal, , 

no abstracts in English

Oral presentation

Development of level 1 PRA methodology for external vessel storage tank of Japan Sodium-cooled Fast Reactor (JSFR) in scheduled refueling

Naruto, Kenichi*; Sugino, Tetsu*; Yamano, Hidemasa; Kurisaka, Kenichi; Nishino, Hiroyuki; Okano, Yasushi

no journal, , 

For SFR, new fuel assembly and used that are stored in EVST having sodium pool. This study identified accident sequence induced core damage by function loss of decay heat removal at scheduled refueling in the EVST and developed methodology for calculation of fuel damage frequency.

Oral presentation

Development of level 1 PRA methodology for external vessel storage tank of Japan Sodium-cooled Fast Reactor (JSFR) in whole-core evacuation

Naruto, Kenichi*; Yamano, Hidemasa; Kurisaka, Kenichi; Nishino, Hiroyuki; Okano, Yasushi

no journal, , 

no abstracts in English

Oral presentation

Level 1 PRA for design works of external vessel storage tank in advanced loop-type sodium-cooled fast reactor

Yamano, Hidemasa; Naruto, Kenichi*; Kurisaka, Kenichi; Nishino, Hiroyuki

no journal, , 

Spent fuels are kept in an external vessel storage tank (EVST) filled with sodium for fuel handling in sodium-cooled fast reactors. This study performed Level 1 PRA for the EVST designed in an advanced loop-type reactor in order to identify dominant accident sequences leading to fuel failure and to quantify fuel damage frequency.

Oral presentation

Level 1 PRA for design works of spent fuel pool in advanced loop-type sodium-cooled fast reactor

Naruto, Kenichi*; Yamano, Hidemasa; Kurisaka, Kenichi; Nishino, Hiroyuki

no journal, , 

no abstracts in English

Oral presentation

Level 1 PRA for external vessel storage tank in whole core refueling in advanced loop-type sodium-cooled fast reactor

Yamano, Hidemasa; Naruto, Kenichi*; Kurisaka, Kenichi; Nishino, Hiroyuki

no journal, , 

Spent fuels are transferred from a reactor core to a spent fuel pool through an external vessel storage tank (EVST) filled with sodium in sodium-cooled fast reactors in Japan. This paper describes identification of dominant accident sequences leading to fuel failure, which was achieved through probabilistic risk assessment for the EVST designed for a next sodium-cooled fast reactor plant system in Japan to improve the EVST design. The safety strategy for the EVST involves whole core refueling-early transfer of all core fuel assemblies into the EVST-assuming a severe situation that results in sodium level reduction leading finally to the top of the reactor core fuel assemblies in a long time. This study introduces the success criteria mitigation along the decay heat decrease over time. Based on the design information, this study has carried out identification of initiating events, event and fault tree analyses, a probability analysis for human error, and quantification of accident sequences. The fuel damage frequency of the EVST was evaluated to be approx. 10$$^{-6}$$/year. This study also quantitatively showed the effectiveness of design improvement.

20 (Records 1-20 displayed on this page)
  • 1