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Onoda, Yuichi; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa
Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 10 Pages, 2024/10
We developed the measures for improving resilience of the sodium-cooled fast reactor structure using the failure mitigation technology and evaluated the effectiveness of the measures. To prevent core damage in the event of an accident progressing to an ultra-high temperature state, both measures to prevent overpressure in the reactor vessel and measures to cool the reactor core are required. As a core cooling measure, we developed a core cooling concept that promotes radiant heat transfer from the reactor vessel and cools the containment vessel outer surface by natural convection named Containment Vessel Auxiliary Cooling System (CVACS). We developed a method to use the reduction rate of core damage frequency as an indicator for effectiveness of the measures for improving resilience. The core damage frequency was evaluated by calculating the core cooling performance using CVACS, reflecting the results of structural analysis and human reliability analysis. By implementing measures for improving resilience in addition to existing measures, the core damage frequency of Japan loop-type sodium-cooled fast reactor caused by LOHRS has been reduced to about one-hundredth of the previous level.
Nishino, Hiroyuki; Kurisaka, Kenichi; Futagami, Satoshi; Watakabe, Tomoyoshi; Yamano, Hidemasa
Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 10 Pages, 2024/10
The reactor vessel (RV) buckling was a dominant contributor to core damage. However, even if the RV is buckled due to seismic shaking, it is expected that the RV maintains stable state without unstable failure such as rupture, collapse. Realistic consideration of the post-buckling behavior is regarded as a measure for improving the resilience in this study. The purpose of this study is to understand the post-buckling deformation behavior of the RV and to evaluate the RV fragility based on fatigue failure. This study performed structural analysis using a finite element method to quantify time histories of displacement, strain, etc. As the result of the analysis, wrinkles of the buckling appeared at the elevation higher than the liquid level in the RV. The largest strain value was also indicated around this elevation. The cumulative fatigue damage fraction was evaluated in this analysis to evaluate the fragility of fatigue failure in addition to the buckling fragility. The result showed that the seismic intensity for the median fragility of the fatigue failure was about six times larger than the design-basis ground motion. This is 1.2 times larger than the buckling-based result, which suggests that realistic evaluation of the post-buckling behavior could contribute to improving the resilience of the nuclear structure.
Nishino, Hiroyuki; Kurisaka, Kenichi; Naruto, Kenichi*; Gondai, Yoji; Yamamoto, Masaya
Mechanical Engineering Journal (Internet), 11(2), p.23-00409_1 - 23-00409_15, 2024/04
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 1E-6 /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; Yamano, Hidemasa; Kurisaka, Kenichi
Mechanical Engineering Journal (Internet), 10(4), p.22-00387_1 - 22-00387_20, 2023/08
For nuclear power plants, probabilistic risk assessment (PRA) should be performed not only against earthquake and tsunami, which are critical events especially in Japan, but also other external hazards such as strong wind. The aim of the present study is to develop a practical PRA methodology for sodium-cooled fast reactors (SFRs) against strong wind, paying attention to the final heat sink, ambient air, that removes decay heat under accident conditions. First, this study used Gumbel distributions to estimate hazard curves of the strong wind based on weather data recorded in Japan. Second, it identified important structures, systems and components (SSCs) for decay heat removal, and developed an event tree that results in core damage, focusing on the impacts of missiles (e.g., steel pipes) caused by strong wind. It also identified missiles that can reach SSCs at elevated places, and calculated the fragility of the SSCs against the missiles as a product of two probabilities. One is a probability of the missiles that would enter an inlet or outlet of the decay heat removal system, and another is a probability of failure caused by missile impacts. Finally, it quantified conditional decay heat removal failure probabilities by introducing the fragilities into the event tree. The core damage frequency (CDF) was estimated at about 5x10-10/y. The dominant sequence is that strong wind causes offsite power loss and missiles, the missiles penetrate the diesel fuel tank, cause a fire, and the fire increases air temperature around the reactor building where air cooler inlets of decay heat removal systems are installed, leads to loss of power for the diesel generator for forced circulation cooling, resulting in loss of decay heat removal. Through the above, this study has developed the practical PRA methodology for SFRs against strong wind.
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.
Kurisaka, Kenichi; Nishino, Hiroyuki; Yamano, Hidemasa
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05
The objective of this study is to develop an effectiveness evaluation methodology of the measures for improving resilience of nuclear structures against excessive earthquake by applying the failure mitigation technology. This study regarded those measures for improving resilience of important structures, systems, and components for safety to enlarge their seismic safety margin. To evaluate effectiveness of those measures, seismic core damage frequency (CDF) is selected as an index. Reduction of CDF as an effectiveness index is quantified by applying seismic PRA technology. Accident sequences leading to loss of decay heat removal are significant contributor to seismic CDF of sodium-cooled fast reactors (SFRs), and those sequences result in core damage via ultra-high temperature condition. This study improved the methodology to evaluate not only the measures against shaking due to excessive earthquake but also the measures at the ultra-high temperature condition. To examine applicability of the improved methodology, a trial calculation was implemented with some assumptions for a loop-type SFR. Within the assumption, the measures for improving resilience were significantly effective for decreasing CDF in excessive earthquake up to several times of a design basis ground motion. Through the applicability examination, the methodology for the effectiveness evaluation was developed successfully.
Kobayashi, Daisuke; Yamamoto, Masahiko; Nishida, Naoki; Miyoshi, Ryuta; Nemoto, Ryo*; Hayashi, Hiroyuki*; Kato, Keisuke; Nishino, Saki; Kuno, Takehiko; Kitao, Takahiko; et al.
Nihon Hozen Gakkai Dai-18-Kai Gakujutsu Koenkai Yoshishu, p.237 - 240, 2022/07
All gloves attached to glove-box in Tokai Reprocessing Plant have a fixed expiration date and have to be replaced every 4-year. However, degrees of glove deterioration are different depending on its usage environment (frequency, chemicals, radiation, etc.), because of rubber products. Therefore, physical properties such as tensile strength, elongation, hardness of gloves are measured, and the life-time of gloves are estimated. As a result, gloves without any defects in its appearance have enough physical property for acceptance criteria of new glove. The extrapolated physical property of glove is sufficiently larger than the reported values of damaged glove. No deterioration in physical properties of gloves, that are periodically replaced without any defects in its appearance, is observed and the usable life-time of the glove is estimated to be 8 years.
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.
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.
Nishino, Hiroyuki; Onoda, Yuichi; Kurisaka, Kenichi; Yamano, Hidemasa
Proceedings of Asian Symposium on Risk Assessment and Management 2021 (ASRAM 2021) (Internet), 10 Pages, 2021/10
The objective of this study is to develop an effectiveness evaluations technology of the measures for improving resilience of nuclear structures against excessive earthquake by introducing the fracture control concept. After analyzing event tree in previous studies of PRA against earthquake, this study identified sequences of protected loss of heat sink and loss of reactor level induced from excessive earthquake as accident sequences in which improving resilience of nuclear structures become effective. This study focused on important components for safety (e.g., reactor vessel, air coolers, pipes of primary loops in decay heat removal systems, etc.) to be used as countermeasures for improving the resilience. Core damage frequency is selected as an index in evaluating effectiveness of the measures for improving the resilience. Seismic safety margin of the components is assumed to be enlarged when the measures for improving the resilience with the fracture control concept are implemented. Through the trial calculation, reduction effect of the core damage frequency was quantified. The result showed that the measures for improving the resilience are significantly effective for decreasing the core damage frequency in excessive earthquake twice higher than a design basis ground motion. The general concept for the effectiveness evaluations technology was formulated.
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%.
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.
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; 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 310/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.
Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi
Mechanical Engineering Journal (Internet), 5(4), p.18-00079_1 - 18-00079_17, 2018/08
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/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.
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.
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.
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
/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.
Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07
