Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 78

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

Development of effectiveness evaluations technology of the measures for improving resilience of nuclear structures at ultra high temperature

Onoda, Yuichi; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

Proceedings of Asian Symposium on Risk Assessment and Management 2021 (ASRAM 2021) (Internet), 11 Pages, 2021/10

The effectiveness evaluations technology of the measures for improving resilience by applying a fracture control concept under ultra-high temperature conditions has developed for prototype sodium-cooled fast reactor Monju as a model plant, and the trial evaluation has conducted using this technology in this paper. The important accident sequences to which the fracture control concept is expected to be applied under ultra-high temperature condition are identified by investigating the results of the existing researches of level-2 probabilistic risk assessment for Monju. Accident sequences categorized in protected loss of heat sink and loss of reactor level are both identified as such important accident sequences which has the potential to prevent core damage. This study has developed the technology to evaluate the effectiveness of improving resilience, where the headings which stand for success or failure of the measures to improve resilience are introduced into the event tree, the branch probability of them is set, and the effectiveness of improving resilience is expressed as the reduction of core damage frequency. As a result of the trial evaluation of the effectiveness for the measures to improve resilience, it is confirmed that core damage frequency can be reduced by applying fracture control concept. The branch probability of the measures to improve resilience proposed in this study is tentatively assigned based on the assumption. This value is expected to be quantified by the forthcoming analyses of the integrity for the reactor vessel structure at ultra-high temperature. The technology developed in this study will be applied for the evaluation of improving resilience of the next generation sodium-cooled fast reactor.

Journal Articles

Internal event level-1 PRA for sodium-cooled fast reactor considering safety measures of defense-in-depth level 1 to 3

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%.

Journal Articles

A Study of probabilistic risk assessment methodology of external hazard combinations; Identification of hazard combination impacts on air-cooling decay heat removal system

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

Proceedings of International Topical Meeting on Probabilistic Safety Assessment and Analysis (PSA 2019), p.274 - 281, 2019/04

A sodium-cooled fast reactor uses the ambient air as an ultimate heat sink to remove decay heat, thus meteorological phenomena can potentially pose risks to the reactor. If a rare and intense external hazard occurs concurrently with another external hazard, it would affect the systems (i.e. air cooler of decay heat removal system). In this study, a new scheme of screening of the external hazard combinations was proposed. The authors classified simultaneous or sequential combinational hazards, and identified associated potential effects in terms of hazard duration and sequential order. As a result, this study identified scenarios of the external hazard combinations of preceding rare and intense external hazard with an following additional external hazard.

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

Development of probabilistic risk assessment methodology against volcanic eruption for sodium-cooled fast reactors

Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamamoto, Takahiro*

ASCE-ASME Journal of Risk and Uncertainty in Engineering Systems, Part B; Mechanical Engineering, 4(3), p.030902_1 - 030902_9, 2018/09

This paper describes volcanic probabilistic risk assessment (PRA) methodology development for sodium-cooled fast reactors. The volcanic ash could potentially clog air filters of air-intakes that are essential for the decay heat removal. The degree of filter clogging can be calculated by atmospheric concentration of ash and tephra fallout duration and also suction flow rate of each component. The atmospheric concentration can be calculated by deposited tephra layer thickness, tephra fallout duration and fallout speed. This study evaluated a volcanic hazard using a combination of tephra fragment size, layer thickness and duration. In this paper, each component functional failure probability was defined as a failure probability of filter replacement obtained by using a grace period to a filter failure limit. Finally, based on an event tree, a core damage frequency was estimated about 3$$times$$10$$^{-6}$$/year in total by multiplying discrete hazard probabilities by conditional decay heat removal failure probabilities. A dominant sequence was led by the loss of decay heat removal system due to the filter clogging after the loss of emergency power supply. In addition, sensitivity analyses have investigated the effects of a tephra arrival reduction factor and pre-filter covering.

Journal Articles

Development of a probabilistic risk assessment methodology against a combination hazard of strong wind and rainfall for sodium-cooled fast reactors

Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi

Mechanical Engineering Journal (Internet), 5(4), p.18-00093_1 - 18-00093_19, 2018/08

This paper describes the development of a probabilistic risk assessment (PRA) methodology against a combination hazard of strong wind and rainfall. In this combination hazard PRA, a hazard curve is evaluated in terms of maximum instantaneous wind speed, hourly rainfall, and rainfall duration. A scenario analysis has provided event sequences resulting from the combination hazard of strong wind and rainfall. The typical event sequence was characterized by the function loss of auxiliary cooling system, of which heat transfer tubes could crack due to cycle fatigue caused by cyclic contacts with rain droplets. This cycle fatigue crack could occur if rain droplets enter into the air cooler of the system following the coolers roof failure due to strong-wind-generated missile impact. This event sequence has been incorporated into an event tree which addresses component failure caused by the combination hazard. As a result, a core damage frequency has been estimated to be about 10$$^{-6}$$/year in total by multiplying discrete hazard frequencies by conditional decay heat removal failure probabilities. The dominant sequence is the manual operation failure of an air cooler damper following the failure of external fuel tank due to the missile impact. The dominant hazard is the maximum instantaneous wind speed of 20-40 m/s, the hourly rainfall of 20-40 mm/h, and the rainfall duration of 0-10 h.

Journal Articles

Development of probabilistic risk assessment methodology of decay heat removal function against combination hazard of low temperature and snow for sodium-cooled fast reactors

Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi

Mechanical Engineering Journal (Internet), 5(4), p.18-00079_1 - 18-00079_17, 2018/08

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

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

Development of probabilistic risk assessment methodology of decay heat removal function against combination hazard of low temperature and snow for sodium-cooled fast reactors

Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07

Journal Articles

Development of probabilistic risk assessment methodology of decay heat removal function against combination hazards of strong wind and rainfall for sodium-cooled fast reactors

Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 12 Pages, 2017/07

This paper describes probabilistic risk assessment (PRA) methodology development against combination hazard of strong wind and rainfall. In this combination hazard PRA, a hazard curve has been evaluated in terms of maximum instantaneous wind speed, hourly rainfall, and rainfall duration. A scenario analysis provided event sequences resulted from the combination hazard of strong wind and rainfall. The event sequence was characterized by the function loss of auxiliary cooling system, of which heat transfer tubes could crack due to cycle fatigue by cyclic contact of rain droplets. This situation could occur if rain droplets ingress into air cooler occurs after the air cooler roof failure due to strong-wind-generated missile impact. This event sequence was incorporated into an event tree which addressed component failure by the combination hazard. Finally, a core damage frequency has been estimated the order of 10$$^{-7}$$/year in total by multiplying discrete hazard frequencies by conditional decay heat removal failure probabilities. A dominant sequence is the failure of the auxiliary cooling system by the missile impact after the failure of external fuel tank by the missile impact. A dominant hazard is the maximum instantaneous wind speed of 40-60 m/s, the hourly rainfall of 20-40 mm/h, and the rainfall duration of 0-10 h.

Journal Articles

Research and development of probabilistic risk assessment methodology for combination event of low temperature and snow

Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi

Nihon Kikai Gakkai Rombunshu (Internet), 83(847), p.16-00392_1 - 16-00392_13, 2017/03

Journal Articles

Development of probabilistic risk assessment methodology against extreme snow for sodium-cooled fast reactor

Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi

Nuclear Engineering and Design, 308, p.86 - 95, 2016/11

 Times Cited Count:4 Percentile:45.9(Nuclear Science & Technology)

This paper describes snow probabilistic risk assessment (PRA) methodology development through external hazard and event sequence evaluations mainly in terms of decay heat removal (DHR) function of a sodium-cooled fast reactor (SFR). Using recent 50-year weather data at a typical Japanese SFR site, snow hazard categories were set for the combination of daily snowfall depth (snowfall speed) and snowfall duration which can be calculated by dividing the snow depth by the snowfall speed. For each snow hazard category, the event sequence was evaluated by event trees which consist of several headings representing the loss of DHR. Snow removal action and manual operation of the air cooler dampers were introduced into the event trees as accident managements. Access route failure probability model was also developed for the quantification of the event tree. In this paper, the snow PRA showed less than 10$$^{-6}$$/reactor-year of core damage frequency. The dominant snow hazard category was the combination of 1-2 m/day of snowfall speed and 0.5-0.75 day of snowfall duration. Importance and sensitivity analyses indicated a high risk contribution to secure the access routes.

Journal Articles

Development of extreme rainfall PRA methodology for sodium-cooled fast reactor

Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 10 Pages, 2016/11

Journal Articles

Event sequence assessment using plant dynamics analysis based on continuous Markov chain process with Monte Carlo sampling assessment of strong wind hazard in sodium cooled fast reactor

Takata, Takashi; Azuma, Emiko*; Nishino, Hiroyuki; Yamano, Hidemasa; Sakai, Takaaki*

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 6 Pages, 2016/11

A new approach has been developed to assess event sequences under external hazard condition 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 strong wind is selected as the external hazard to assess the plant safety in a loop type sodium cooled fast reactor. As a result, it is demonstrated that the plant state is quite safe in case of the strong wind because multiple failures of the air coolers in the auxiliary cooling system (ACS) has a quite low probability. Furthermore, a weight factor is introduced so as to investigate the low failure probability events with a comparative small number of the sampling.

Journal Articles

Development of risk assessment methodology against natural external hazards for sodium-cooled fast reactors; Project overview and margin assessment methodology against volcanic eruption

Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi; Okano, Yasushi; Sakai, Takaaki; Yamamoto, Takahiro*; Ishizuka, Yoshihiro*; Geshi, Nobuo*; Furukawa, Ryuta*; Nanayama, Futoshi*; et al.

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 12 Pages, 2016/10

This paper describes mainly volcanic margin assessment methodology development in addition to the project overview. The volcanic tephra could potentially clog filters of air-intakes that need the decay heat removal. The filter clogging can be calculated by atmospheric concentration and fallout duration of the volcanic tephra and also suction flow rate of each component. In this paper, the margin was defined as a grace period to a filter failure limit. Consideration is needed only when the grace period is shorter than the fallout duration. The margin by component was calculated using the filter failure limit and the suction flow rate of each component. The margin by sequence was evaluated based on an event tree and the margin by component. An accident management strategy was also suggested to extend the margin; for instance, manual trip of the forced circulation operation, sequential operation of three air coolers, and covering with pre-filter.

Journal Articles

Research and development of probabilistic risk assessment methodology for combination event of low temperature and snow

Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi

Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06

The objective of this study is to develop probabilistic risk assessment (PRA) methodology for combination event of low temperature and snow by focusing attention on decay heat removal system (DHRS) of sodium-cooled fast reactor. For this combination event, annual excess probability depending on the hazard intensity was statistically estimated based on the meteorological data. Event tree was developed by considering the impact of low temperature and snow on DHRS: e.g., plug at the air intake of ultimate heat sink and of emergency diesel generator due to accumulated snow, failure of air intake filter due to deposited snow, possibility of freezing of cooling circuits. Recovery actions (i.e., snow removal and filter replacement) were considered in the event tree. Quantification of the event tree showed that dominant core damage sequence is loss of access route for snow removal against the combination event at daily snowfall of 3m/day continued during 24h.

Journal Articles

Development of risk assessment methodology of decay heat removal function against natural external hazards for sodium-cooled fast reactors; Project overview and volcanic PRA methodology

Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi; Okano, Yasushi; Sakai, Takaaki; Yamamoto, Takahiro*; Ishizuka, Yoshihiro*; Geshi, Nobuo*; Furukawa, Ryuta*; Nanayama, Futoshi*; et al.

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 10 Pages, 2016/06

This paper describes mainly volcanic probabilistic risk assessment (PRA) methodology development for sodium-cooled fast reactors in addition to the project overview. The volcanic ash could potentially clog air filters of air-intakes that are essential for the decay heat removal. The degree of filter clogging can be calculated by atmospheric concentration of ash and tephra fallout duration and also suction flow rate of each component. The atmospheric concentration can be calculated by deposited tephra layer thickness, tephra fallout duration and fallout speed. This study evaluated a volcanic hazard using a combination of tephra fragment size, layer thickness and duration. In this paper, each component functional failure probability was defined as a failure probability of filter replacement obtained by using a grace period to a filter failure limit. Finally, based on an event tree, a core damage frequency was estimated about 3$$times$$10$$^{-6}$$/year in total by multiplying discrete hazard probabilities by conditional decay heat removal failure probabilities. A dominant sequence was led by the loss of decay heat removal system due to the filter clogging after the loss of emergency power supply. A dominant volcanic hazard was 10$$^{-2}$$ kg/m$$^{3}$$ of atmospheric concentration, 0.1 mm of tephra diameter, 50-75 cm of deposited tephra layer thickness, and 1-10 hr of tephra fallout duration.

Journal Articles

Development of risk assessment methodology against external hazards for sodium-cooled fast reactors

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.

Journal Articles

Development of risk assessment methodology against natural external hazards for sodium-cooled fast reactors; Project overview and strong wind PRA methodology

Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi; Okano, Yasushi; Sakai, Takaaki; Yamamoto, Takahiro*; Ishizuka, Yoshihiro*; Geshi, Nobuo*; Furukawa, Ryuta*; Nanayama, Futoshi*; et al.

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.454 - 465, 2015/05

This paper describes mainly strong wind PRA methodology development in addition to the project overview. In developing the strong wind PRA methodology, hazard curves were estimated by using Weibull and Gumbel distributions based on weather data recorded in Japan. The obtained hazard curves were divided into five discrete categories for event tree quantification. Next, failure probabilities for decay heat removal related components were calculated as a product of two probabilities: i.e., a probability for the missiles to enter the intake or outtake in the decay heat removal system, and fragility caused by the missile impacts. Finally, based on the event tree, the core damage frequency was estimated about 6$$times$$10$$^{-9}$$/year by multiplying the discrete hazard probabilities in the Gumbel distribution by the conditional decay heat removal failure probabilities. A dominant sequence was led by the assumption that the operators could not extinguish fuel tank fire caused by the missile impacts and the fire induced loss of the decay heat removal system.

78 (Records 1-20 displayed on this page)