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Journal Articles

For better understanding of PRA; Guidance for better usage and application of PRA, 3; Consideration of external hazard in PRA

Takata, Takashi*; Yamano, Hidemasa; Narumiya, Yoshiyuki*

Nippon Genshiryoku Gakkai-Shi, 62(8), p.448 - 451, 2020/08

This report describes risk analysis applicable to selection and evaluation of external hazards in risk assessment. Volcanic ash hazard evaluation is shown as one of hazard evaluations. The objective of the risk assessment is to secure and improve the safety of nuclear installation. This report also describes discussion on the process of response informed based on risk evaluation against external hazards.

JAEA Reports

Development of technology to prevent scattering of radioactive materials in fuel debris retrieval (Contract research); FY2018 Center of World Intelligence Project for Nuclear Science/Technology and Human Resource Development

Collaborative Laboratories for Advanced Decommissioning Science; The University of Tokyo*

JAEA-Review 2019-037, 90 Pages, 2020/03

JAEA-Review-2019-037.pdf:7.0MB

JAEA/CLADS, had been conducting the Center of World Intelligence Project for Nuclear Science/Technology and Human Resource Development (hereafter referred to "the Project") in FY2018. The Project aims to contribute to solving problems in nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2018, this report summarizes the research results of the "Development of Technology to Prevent Scattering of Radioactive Materials in Fuel Debris Retrieval". The objective of the present study is to clarify the behavior of microparticles in gas and liquid phases in order to steadily confine radioactive microparticles at the time of debris retrieval in Fukushima Daiichi Nuclear Power Station. In addition, as measures to prevent scattering, we will evaluate and develop methods by experiments and simulation as to; (1) a method to suppress the scattering with minimum amount of water utilizing water spray etc., and (2) a method to suppress the scattering by solidifying fuel debris.

Journal Articles

Evaluation of the effects of differences in building models on the seismic response of a nuclear power plant structure

Choi, B.; Nishida, Akemi; Muramatsu, Ken*; Takada, Tsuyoshi*

Nippon Jishin Kogakkai Rombunshu (Internet), 20(2), p.2_1 - 2_16, 2020/02

no abstracts in English

Journal Articles

Development of a fast reactor and related thermal hydraulics studies in Japan

Ohshima, Hiroyuki; Kamide, Hideki

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.2095 - 2107, 2019/08

Development of a sodium-cooled fast reactor has been implemented in Japan from the viewpoint of severe accident countermeasures. This paper describes the progress of research and development related to safety enhancement and the severe accident countermeasures. A volcanic PRA methodology was developed for the proper consideration of external hazards. Water and sodium experiments were carried out for the decay heat removal in a core disruptive accident (CDA), and also thermal hydraulic interactions between the core and upper plenum where dipped heat exchanger was operated. In order to elucidate the behavior of molten fuel during CDA, basic experiments of core melt fragmentation in deep and shallow sodium pools were carried out. X-ray visualization showed the liquid column of molten steel was intensively fragmented nearly simultaneously with a rapid expansion of sodium vapor.

Journal Articles

Visualizing an ignition process of hydrogen jets containing sodium mist by high-speed imaging

Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya*; Uno, Masayoshi*

Journal of Nuclear Science and Technology, 56(6), p.521 - 532, 2019/06

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

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

Epistemic Uncertainty Quantification of Floor Responses for a Nuclear Reactor Building

Choi, B.; Nishida, Akemi; Li, Y.; Muramatsu, Ken*; Takada, Tsuyoshi*

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

After the 2011 Fukushima accident, nuclear power plants are required to take countermeasures against accidents beyond design basis conditions. In seismic probabilistic risk assessment (SPRA), uncertainty can be classified as either aleatory uncertainty, which cannot be reduced, or epistemic uncertainty, which can be reduced with additional knowledge and/or information. To improve the reliability of SPRA, efforts should be made to identify and reduce the epistemic uncertainty caused by the lack of knowledge. In this study, we focused on the difference in seismic response by modeling methods, which is related epistemic uncertainty. We conducted a seismic response analysis with two kinds of modeling methods; a three-dimensional finite-element model and a conventional sway-rocking stick model, by using simulated various input ground motions, which is related to aleatory uncertainty. And then we quantified the seismic floor response results of the various input ground motions of each modeling methods. For the uncertainty quantification related to different modeling methods, we further perform a statistical analysis of the floor response results of the nuclear reactor building. Finally, we discussed how to utilize the results from these calculations for the quantification of uncertainty in fragility analysis for SPRA.

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

Uncertainty evaluation of seismic response of a nuclear facility using simulated input ground motions

Choi, B.; Nishida, Akemi; Muramatsu, Ken*; Takada, Tsuyoshi*

Proceedings of 12th International Conference on Structural Safety & Reliability (ICOSSAR 2017) (USB Flash Drive), p.2206 - 2213, 2017/08

In order to clarify the influence of the difference of modeling method on the variation of the result of seismic response analysis of nuclear facility, seismic response analysis using various simulated input ground motions was carried out and the sensitivity analyses of the variations in seismic response was conducted. In particular, we focused on the maximum acceleration response of reactor building shear walls, the effect of modeling method on response result and the factors of response variation were described and discussed.

Journal Articles

Uncertainty assessment of structural modeling in the seismic response analysis of nuclear facilities

Choi, B.; Nishida, Akemi; Muramatsu, Ken*; Takada, Tsuyoshi*

Transactions of 24th International Conference on Structural Mechanics in Reactor Technology (SMiRT-24) (USB Flash Drive), 10 Pages, 2017/08

In order to clarify the influence of the modeling method on the result of seismic response analysis of nuclear facility, seismic response analysis using various simulated input ground motions was carried out and the uncertainty of response results were statistically analyzed. In particular, we focused on the difference of the response due to the structural modeling method (a conventional sway-rocking model and 3D FE model), and the relations among the input level, floor position, and response results were described and discussed.

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

Development of margin assessment methodology of decay heat removal function against external hazards

Yamano, Hidemasa

Genshiryoku Shisutemu Kenkyu Kaihatsu Oyobi Eichi O Kesshushita Genshiryoku Kagaku Gijutsu, Jinzai Ikusei Suishin Jigyo Heisei-28-Nendo Seika Hokokukai Shiryoshu (Internet), 4 Pages, 2017/01

This study has carried out R&D of margin assessment methodology of decay heat removal function for sodium-cooled fast reactors against external hazards in the MEXT innovative nuclear system R&D project from 2012 to 2016. In 2014, the development has been completed for hazard evaluation and event sequence evaluation methods against combination hazards of snow and low temperature as well as strong wind and rainfall. In forest fire, a combination hazard evaluation methodology has been developed by taking into account lightning.

Journal Articles

Event sequence analyses of a forest fire heat effect on a sodium-cooled fast reactor for an external hazard PRA methodology development

Okano, Yasushi; Yamano, Hidemasa

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 9 Pages, 2016/11

This paper deals with an event sequence by forest fire heat effect on a decay heat removal function of a sodium-cooled fast reactor. Related to the potential vulnerability, an event scenario was developed using conservative assumptions. An event tree was developed with an initiating event of the loss of off-site power, and the headings are related to "external diesel fuel tanks", "emergency diesel generator and its auxiliary system", "alternative cooling system and its power source", and "decay heat air cooler". A failure probability on each heading was given from a fragility curve as a function of reaction intensity or by assumptions based on conservative models. A core damage frequency, under the conditional of the loss of off-site power, was conservatively evaluated around 10$$^{-7}$$/year. A key heading in the event tree with large effect on the frequency is the intactness of the external diesel fuel tanks.

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

Dynamic and interactive approach to level 2 PRA using continuous Markov process with Monte Carlo Method

Jang, S.*; Yamaguchi, Akira*; Takata, Takashi

Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 11 Pages, 2016/10

The current approach to Level 2 probabilistic risk assessment (PRA) using the conventional event-tree (ET)/fault-tree (FT) methodology requires pre-specifications of event order occurrence and component failure probabilities which may vary significantly in the presence of uncertainties. In the present study, a new methodology is proposed to quantify the level 2 PRA in which the accident progression scenarios are dynamic and interactive with the instantaneous plant state and related phenomena. The accident progression is treated as a continuous Markov process and the transition probabilities are evaluated based on the computation of plant system thermal-hydraulic dynamics. A Monte Carlo method is used to obtain the resultant probability of the radioactive material release scenarios. The methodology is applied to the protected loss of heat sink accident scenario of the level 2 PRA of a generation IV fast reactor.

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

DNA gyrase of ${it Deinococcus radiodurans}$ is characterized as Type II bacterial topoisomerase and its activity is differentially regulated by PprA in vitro

Kota, S.*; Rajpurohit, Y. S.*; Charaka, V. K.*; Sato, Katsuya; Narumi, Issey*; Misra, H, S.*

Extremophiles, 20(2), p.195 - 205, 2016/03

 Times Cited Count:8 Percentile:53.09(Biochemistry & Molecular Biology)

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