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Nishino, Hiroyuki; Kurisaka, Kenichi; Naruto, Kenichi*; Gondai, Yoji; Yamamoto, Masaya
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
The effectiveness evaluation of safety measures against severe accident is necessary for restart of experimental sodium-cooled fast reactor "Joyo" in Japan. These safety measures correspond to those in defense-in-depth (DiD) level 4. In the previous study, a level-1 probabilistic risk assessment (PRA) at power was performed to calculate frequencies of the accident sequences of failure of safety measures in DiD level 1 to 3, to identify dominant accident sequence groups, and to identify dominant accident sequence for selecting important accident sequences in each accident sequence group which are needed for implementing the effectiveness evaluation of safety measures in DiD level 4. Based on this, the present study implemented level-1 PRA at power to show quantitatively reduction of those occurrence frequency by the safety measure in the DiD level 4. As the result, the frequency of each accident sequence group decreased significantly, and total frequency of the accident sequence groups decreased to about 1
10
/reactor-year which is about 1/1000 times the one estimated in the previous study. The protected loss of heat sink was the largest contributor in all the accident groups and a dominant accident sequence in each accident group was also identified in this study.
Nishino, Hiroyuki; Kurisaka, Kenichi; Naruto, Kenichi*; Gondai, Yoji; Yamamoto, Masaya; Yamano, Hidemasa
Proceedings of Asian Symposium on Risk Assessment and Management 2020 (ASRAM 2020) (Internet), 12 Pages, 2020/11
The objective of this study is to evaluate the occurrence frequency of accident sequences which may lead to core damage if provisions in defense in depth (DiD) level 1 to 3 are the only safety measures. For this objective, the existing safety measures in this SFR are categorized into those for the DiD level 1-3 and those for the DiD level 4. The safety measures for the DiD level 1-3 are as follows; (1) main reactor shutdown system, (2) double boundary structure in the primary main and auxiliary cooling system and the reactor vessel, which maintain the reactor coolant level sufficient for coolant circulation in the primary main cooling system, (3) decay heat removal in a forced circulation mode. Accident sequences are categorized into typical SFR-specific groups and station blackout (SBO) in this study. The SFR-specific groups are unprotected loss of flow, unprotected transient over power, unprotected loss of heat sink, loss of reactor level, and protected loss of heat sink (PLOHS). The occurrence frequency of these accident sequence groups was quantified to identify major contributors. As the result, PLOHS excluding SBO was indicated as the dominant contribution of 80% or more in the all accident sequence groups and the annual occurrence frequency of the PLOHS was 1.0E-4 order of magnitude. For the PLOHS, loss of offsite power (LOOP) was indicated as major contribution of 30% in initiating events. In the accident sequences of the PLOHS initiated from LOOP, a dominant sequence was combination of common cause failure of primary pumps in the main cooling system and failure-to-start of the auxiliary cooling system after LOOP. The second dominant contribution (15% or more) in the all accident sequence groups is PLOHS in SBO (i.e., decay heat removal failure due to SBO). Each of the other accident sequence groups was 1%.
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
/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.
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 /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.
Nishimura, Masahiro; Fukano, Yoshitaka; Kurisaka, Kenichi; Naruto, Kenichi*
Journal of Nuclear Science and Technology, 54(11), p.1178 - 1189, 2017/11
Times Cited Count：3 Percentile：28.82(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.
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 /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.
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.
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：25.85(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.
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
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.
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
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.
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.
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.
Naruto, Kenichi*; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa; Okano, Yasushi; Okamura, Shigeki*; Eto, Masao*
no journal, ,
no abstracts in English
Naruto, Kenichi*; Sugino, Tetsu*; Yamano, Hidemasa; Kurisaka, Kenichi; Nishino, Hiroyuki; Okano, Yasushi
no journal, ,
no abstracts in English
Miyabe, Takaaki*; Naruto, Kenichi*; Sugino, Tetsu*; Yamano, Hidemasa; Kurisaka, Kenichi; Nishino, Hiroyuki; Okano, Yasushi
no journal, ,
no abstracts in English
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.
Naruto, Kenichi*; Yamano, Hidemasa; Kurisaka, Kenichi; Nishino, Hiroyuki; Okano, Yasushi
no journal, ,
no abstracts in English
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.
