Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Iwasawa, Yuzuru; Matsumoto, Toshinori; Moriyama, Kiyofumi*
JAEA-Data/Code 2025-001, 199 Pages, 2025/06
JAEA-Data-Code-2025-001.pdf:9.71MB
A steam explosion is defined as a phenomenon that occurs when a hot liquid comes into contact with a low-temperature cold liquid with volatile properties. The rapid transfer of heat from the hot liquid to the cold liquid results in a chain reaction of the explosive vaporization of the cold liquid and fine fragmentation of the hot liquid. The explosive vaporization of the cold liquid initiates the propagation of shock waves in the cold liquid. The expansion of the hot and cold liquid mixture exerts mechanical forces on the surrounding structures. In severe accidents of light water reactors, the molten core (melt) is displaced into the coolant water, resulting in fuel-coolant interactions (FCIs). The explosive FCI, referred to as a steam explosion, has been identified as a significant safety assessment issue as it can compromise the integrity of the primary containment vessel. The JASMINE code is an analytical tool developed to evaluate the mechanical forces imposed by steam explosions in nuclear reactors. It performs numerical simulations of steam explosions in a mechanistic manner. The present report describes modeling concepts, basic equations, numerical solutions, and example simulations, as well as instructions for input preparation, code execution, and the use of supporting tools for practical purpose. The present report is the updated version of the "Steam Explosion Simulation Code JASMINE v.3 User's Guide, JAEA-Data/ Code 2008-014". The present report was compiled and updated based on the latest version of the code, JASMINE 3.3c, with corrections for minor errors of the distributed code JASMINE 3.3b, and conformance to recently widely used compilers on UNIX-like environments such as the GNU compiler. The numerical simulations described in the present report were obtained using the latest version JASMINE 3.3c. The latest parameter adjustment method for a model parameter proposed by the previous study is employed to conduct the numerical simulations.
Zheng, X.; Tamaki, Hitoshi; Shibamoto, Yasuteru; Maruyama, Yu
Nihon Genshiryoku Gakkai-Shi ATOMO
, 66(11), p.565 - 569, 2024/11
no abstracts in English
Kurisaka, Kenichi; Nishino, Hiroyuki; Yamano, Hidemasa
Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 10 Pages, 2024/10
The objective of this study is to implement an effectiveness evaluation of the measures for improving resilience of nuclear structures against excessive earthquake. In this study, those measures for improving resilience have an effect 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. Target system is a loop-type next-generation sodium-cooled fast reactor, which adopts the building isolated from horizontal seismic ground motion. Even if the reactor vessel (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 resilience in this study. We set two cases for improving the resilience in the accident sequences analysis. The first case assumes low-cycle fatigue failure after buckling as the realistic failure mode of the RV, and we applied the fragility evaluated in our study. After the RV fatigue failure, the behavior of failure propagation is very uncertain. As the second case, the median seismic capacity to loss of reactor level is assumed to be slightly larger than that of fatigue failure of the RV. Analyses for both cases were performed, and the results were compared to the base case indicating significant reduction of CDF. Within the assumption, the measures for improving the resilience were significantly effective for decreasing CDF in excessive earthquake up to several times of a design basis ground motion. The seismic PRA technology could serve to the effectiveness evaluation of the measures for improving resilience of nuclear structures against excessive earthquake.
Collaborative Laboratories for Advanced Decommissioning Science; Tokyo Institute of Technology*
JAEA-Review 2024-012, 122 Pages, 2024/09
JAEA-Review-2024-012.pdf:6.31MB
The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2022. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station (hereafter referred to "1F"), Tokyo Electric Power Company Holdings, Inc. (TEPCO). 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 FY2021, this report summarizes the research results of the "Challenge of novel hybrid-waste-solidification of mobile nuclei generated in Fukushima Nuclear Power Station and establishment of rational disposal concept and its safety assessment" conducted in FY2022. The present study aims to establish the rational waste disposal concept of a variety of wastes generated in 1F based on the hybrid-waste-solidification by the Hot Isostatic Press (HIP) method. The ceramics form with target elements, mainly iodine, which is difficult to immobilize, and Minor actinides such as Am, an alphaemitter and heat source, are HIPed with well-studied materials such as SUS and zircaloy, which make the long-term stability evaluation and safety assessment possible.
Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi
Journal of Nuclear Science and Technology, 61(7), p.935 - 957, 2024/07
Times Cited Count：2 Percentile：46.61(Nuclear Science & Technology)Kurisaka, Kenichi
Mechanical Engineering Journal (Internet), 11(2), p.23-00377_1 - 23-00377_14, 2024/04
This study aims to understand the time-dependent change in the occurrence rate of leak from steam generator (SG) tubes in sodium-cooled fast reactors (SFRs). The target SFRs in the present paper are Phenix in France and BN-600 in Russia. By reviewing publicly available literature that show data from the SFRs, we have investigated the numbers of tube-to-tubeplate welds and tube-to-tube welds, heat transfer areas of tube base metal, operating hours of SGs, dates when SG tube leak occurred, locations of leak, and corrective actions taken after tube leak events, such as replacement of the module, in which a leak occurred. Based on these, we have estimated the time to leak and quantitatively analyzed the time-dependent change of the occurrence rates of SG tube leak for each of the above-mentioned parts by hazard plotting method. The results show that the rates of both Phenix and BN-600 decreased over time. For Phenix, this is probably thanks to improved welding and SG operating conditions. For BN-600, it seems that in many cases, the probable cause of the leak was initial defects that developed to failure during the early stage of reactor operation, and that no special countermeasure was taken in the later stages. Therefore, it would be natural to assume that the rate simply decreased over time. The rate of leak at tube-to-tube welds in Phenix shows significant increase in a short term after a certain period of time. This can be caused by thermal stress repeatedly exerted on the materials.
Watanabe, So; Takahatake, Yoko; Hasegawa, Kenta; Goto, Ichiro*; Miyazaki, Yasunori; Watanabe, Masayuki; Sano, Yuichi; Takeuchi, Masayuki
Mechanical Engineering Journal (Internet), 11(2), p.23-00461_1 - 23-00461_10, 2024/04
Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Proceedings of the Institution of Mechanical Engineers, Part O; Journal of Risk and Reliability, 237(5), p.947 - 957, 2023/10
Times Cited Count：5 Percentile：55.03(Engineering, Multidisciplinary)Probabilistic risk assessment (PRA) is a method used to assess the risks associated with large and complex systems. However, the timing at which nuclear power plant structures, systems, and components are damaged is difficult to estimate if the risk of an external event is evaluated using conventional PRA based on event trees and fault trees. A methodology coupling thermal-hydraulic analysis with external event simulations using Risk Assessment with Plant Interactive Dynamics (RAPID) is therefore proposed to overcome this limitation. A flood propagation model based on Bernoulli's theorem was applied to represent internal flooding in the turbine building of the pressurized water reactor. Uncertainties were also taken into account, including the flow rate of the floodwater source and the failure criteria for the mitigation systems. The simulated recovery actions included the operator isolating the floodwater source and using a drainage pump; these actions were modeled using several simplifications. Overall, the results indicate that combining isolation and drainage can reduce the conditional core damage probability upon the occurrence of flooding by approximately 90%.
Kubo, Kotaro; Tanaka, Yoichi; Hakuta, Yuto*; Arake, Daisuke*; Uchiyama, Tomoaki*; Muramatsu, Ken
Mechanical Engineering Journal (Internet), 10(4), p.23-00051_1 - 23-00051_17, 2023/08
The significance of probabilistic risk assessments (PRAs) of nuclear power plants against external events was re-recognized after the Fukushima Daiichi Nuclear Power Plant accident. Regarding the seismic PRA, handling correlated failures of systems, components, and structures (SSCs) is very important because this type of failure negatively affects the redundancy of accident mitigation systems. The Japan Atomic Energy Research Institute initially developed a fault tree quantification methodology named the direct quantification of fault tree using Monte Carlo simulation (DQFM) to handle SSCs' correlated failures in detail and realistically. This methodology allows quantifying the top event occurrence probability by considering correlated uncertainties related to seismic responses and capacities with Monte Carlo sampling. The usefulness of DQFM has already been demonstrated. However, improving its computational efficiency would allow risk analysts to perform several analyses. Therefore, we applied quasi-Monte Carlo and importance sampling to the DQFM calculation of simplified seismic PRA and examined their effects. Specifically, the conditional core damage probability of a hypothetical pressurized water reactor was analyzed with some assumptions. Applying the quasi-Monte Carlo sampling accelerates the convergence of results at intermediate and high ground motion levels by an order of magnitude over Monte Carlo sampling. The application of importance sampling allows us to obtain a statistically significant result at a low ground motion level, which cannot be obtained through Monte Carlo and quasi-Monte Carlo sampling. These results indicate that these applications provide a notable acceleration of computation and raise the potential for the practical use of DQFM in risk-informed decision-making.
Minari, Eriko*; Kabasawa, Satsuki; Mihara, Morihiro; Makino, Hitoshi; Asano, Hidekazu*; Nakase, Masahiko*; Takeshita, Kenji*
Journal of Nuclear Science and Technology, 60(7), p.793 - 803, 2023/07
Times Cited Count：3 Percentile：42.88(Nuclear Science & Technology)Kubo, Kotaro
Science and Technology of Nuclear Installations, 2023, p.7402217_1 - 7402217_12, 2023/06
Times Cited Count：2 Percentile：46.61(Nuclear Science & Technology)Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
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.
Kurisaka, Kenichi
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
This study aims to understand a time trend of the occurrence rate of steam generator (SG) tube leak in the existing sodium-cooled fast reactors (SFRs) based on the observed data. The target on SFRs in the present paper is Phenix in France and BN600 in Russia. From the open literature review, we investigated the number of tube-to-tube plate weld, the number of tube-to-tube weld, heat transfer area of tube base metal, operating time of SGs, dates when SG tube leak occurred, leaked location, corrective action after tube leak such as replacement of leaked module. Based on these observed data, time to leak is estimated and then time trend of the occurrence rate of SG tube leak for each of the above-mentioned parts was quantitatively analyzed by the hazard plotting method. As a result, the rate of leak at tube-to-tube weld in Phenix shows increase with time due to probable cause of cyclic thermal stress in a short term. As for a long-term trend, the rate of tube leak in both Phenix and BN600 SGs indicated decrease with time probably thanks to improvement in welding and in SG operating condition and to removal of initial failure.
Kubo, Kotaro; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Journal of Nuclear Science and Technology, 60(4), p.359 - 373, 2023/04
Times Cited Count：9 Percentile：82.35(Nuclear Science & Technology)Probabilistic risk assessment (PRA) is an essential approach to improving the safety of nuclear power plants. However, this method includes certain difficulties, such as modeling of combinations of multiple hazards. Seismic-induced flooding scenario includes several core damage sequences, i.e., core damage caused by earthquake, flooding, and combination of earthquake and flooding. The flooding fragility is time-dependent as the flooding water propagates from the water source such as a tank to compartments. Therefore, dynamic PRA should be used to perform a realistic risk analysis and quantification. This study analyzed the risk of seismic-induced flooding events by coupling seismic, flooding, and thermal-hydraulics simulations, considering the dependency between multiple hazards explicitly. For requirements of safety improvement, especially in light of the Fukushima Daiichi Nuclear Power Plant accident, sensitivity analysis was performed on the seismic capacity of systems, and the effectiveness of alternative steam generator injection by a portable pump was estimated. We demonstrate the use of this simulation-based dynamic PRA methodology to evaluate the risk induced by a combination of hazards.
Collaborative Laboratories for Advanced Decommissioning Science; Tokyo Institute of Technology*
JAEA-Review 2022-072, 116 Pages, 2023/03
JAEA-Review-2022-072.pdf:6.32MB
The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2021. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station (1F), Tokyo Electric Power Company Holdings, Inc. (TEPCO). 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 FY2021, this report summarizes the research results of the "Challenge of novel hybrid-waste-solidification of mobile nuclei generated in Fukushima Nuclear Power Station and establishment of rational disposal concept and its safety assessment" conducted in FY2021. The present study aims to establish the rational waste disposal concept of a variety of wastes generated in 1F by the novel hybrid-waste-solidification. The phosphate form of ALPS sediment wastes containing Eu
, Ce
, Sr
and Cs
were synthesized as well as radioactive 
Sr, 
Cs and 
I which are both 
emitters, AREVA sludge and Iodine Calcium apatite were synthesized, and they were processed to the stabilization treatment such as sintering and Spark Plasma ...
Togawa, Orihiko; Okura, Takehisa; Kimura, Masanori
JAEA-Review 2022-049, 76 Pages, 2023/01
JAEA-Review-2022-049.pdf:3.74MB
Before construction and after operation of nuclear facilities, environmental consequence assessments are conducted for normal operation and an emergency. These assessments mainly aim at confirming safety for the public around the facilities and producing relief for them. Environmental consequence assessments are carried out using observations/ measurements by environmental monitoring and/or model predictions by calculation models, sometimes using either of which and at other times using both them, according to the situations and necessities. First, this report investigates methods, roles, merits/demerits and relationship between observations/measurements and model predictions which are used for environmental consequence assessments of nuclear facilities, especially holding up a spent nuclear fuel reprocessing plant at Rokkasho, Aomori as an example. Next, it explains representative examples of utilization of data on observations/measurements and results on model predictions, and considers points of attention at using them. Finally, the report describes future direction, for example, improvements of observations/measurements and model predictions, and fusion of both them.
Kubo, Kotaro; Tanaka, Yoichi*; Ishikawa, Jun
Proceedings of the Institution of Mechanical Engineers, Part O; Journal of Risk and Reliability, 11 Pages, 2023/00
Times Cited Count：1 Percentile：21.45(Engineering, Multidisciplinary)Nakamura, Hideo; Bentaib, A.*; Herranz, L. E.*; Ruyer, P.*; Mascari, F.*; Jacquemain, D.*; Adorni, M.*
Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 10 Pages, 2022/10
Yamaguchi, Akinori*; Yokotsuka, Muneyuki*; Furuta, Masayo*; Kubota, Kazuo*; Fujine, Sachio*; Mori, Kenji*; Yoshida, Naoki; Amano, Yuki; Abe, Hitoshi
Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 21(4), p.173 - 182, 2022/09
Risk information obtained from probabilistic risk assessment (PRA) can be used to evaluate the effectiveness of measures against severe accidents in nuclear facilities. The PRA methods used for reprocessing facilities are considered immature compared to those for nuclear power plants, and to make the methods mature, reducing the uncertainty of accident scenarios becomes crucial. In this paper, we summarized the results of literature survey on the event progression of evaporation to dryness caused by boiling of high-level liquid waste (HLLW) which is a severe accident in reprocessing facilities and migration behavior of associated radioactive materials. Since one of the important characteristics of Ru is its tendency to form volatile compounds over the course of the event progression, the migration behavior of Ru is categorized into four stages based on temperature. Although no Ru has been released in the waste in the high temperature region, other volatile elements such as Cs could be released. Sufficient experimental data, however, have not been obtained yet. It is, therefore, necessary to further clarify the migration behavior of radioactive materials that predominantly depends on temperature in this region.
