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Kamiya, Tomohiro; Ono, Ayako; Tada, Kenichi; Akie, Hiroshi; Nagaya, Yasunobu; Yoshida, Hiroyuki; Kawanishi, Tomohiro
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 8 Pages, 2022/11
JAEA started to develop the advanced reactor analysis code JAMPAN (JAEA advanced multi-physics analysis platform for nuclear systems). The current version of JAMPAN handles the continuous energy Monte Carlo code MVP and the detailed thermal-hydraulics analysis code for multiphase and multicomponent JUPITER. JAMPAN is designed to consider the extensibility and it does not depend on the analysis codes. All calculations in JAMAPAN are not directly connected. JAMPAN has data containers, and all input and output data of each analysis code are set in these data containers. JAMPAN will easily exchange the calculation codes and add the other calculations, e.g., structure calculation and irradiation calculation since the input and the output format of each code has no impact on the other calculation codes. The 4 by 4 pin-cell geometry was used as the demonstration calculation of JAMPAN and the physically reasonable calculation results were obtained.
Yoshida, Hiroyuki; Horiguchi, Naoki; Ono, Ayako; Furuichi, Hajime*; Katono, Kenichi*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 7 Pages, 2022/08
Doi, Daisuke
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 7 Pages, 2022/08
Sun, G.*; Zhan, Y.*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Okano, Yasushi
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08
When a liquid sodium leakage accident occurs in a sodium-cooled fast reactor, the injected sodium collides with structures to produce splashing droplets, which can result in a violent combustion. According to previous studies on circular nozzles, the amount of splash is affected by the state of the jet at the moment of impact. However, the outlet shape of damaged area is hardly to be circular; and meanwhile it influences the flow pattern of jet a lot. Considering about this, in the present work, high-speed cameras were used to observe the jet discharged from oval nozzles vertically downward to investigate the falling process of the jet. The result shows that surface wave appears on the jet and within a certain range of flow velocity it can be observed obviously, meanwhile accelerate the breakup of jet.
Okajima, Satoshi; Mori, Takero; Kikuchi, Norihiro; Tanaka, Masaaki; Miyazaki, Masashi
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 7 Pages, 2022/08
In this paper, we propose the simplified procedure to estimate failure probability of components subjected to thermal transient for the design optimization. Failure probability can be commonly used as an indicator of component integrity for various failure mechanisms. In order to reduce number of analyses required for one estimation, we have adopted the First Order Second Moment (FOSM) method as the estimation method of failure probability on the process of the optimization, and an orthogonal table in experiment design method is utilized to define conditions of the analyses for the evaluation of the input parameters for the FOSM method. Through the demonstration study to optimize thickness of cylindrical vessel subjected to thermal transient derived from shutdown, we confirmed that the procedure can evaluate the failure probability depending on the cylinder thickness with practical calculation cost.
Aoki, Takeshi; Sato, Hiroyuki
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08
High temperature gas-cooled reactor (HTGR) has a potential to produce competitive and large amount of carbon-free hydrogen. It is required to establish the control method and system for the HTGR hydrogen production system to maintain its normal operation against the abnormality in the hydrogen production facility through performance evaluations of the control system by transient thermal-hydraulic analysis. In the present study, the reactor response against the disturbance in the reactor inlet coolant temperature was revealed in the HTGR hydrogen production system. The analytical results showed that the reactor outlet coolant control system enabled to control the variation of the reactor outlet coolant temperature was less than 4
C against 30C of large disturbance in the reactor inlet coolant temperature and to maintain its normal operation in the HTGR hydrogen production system. Thus, the effectiveness of the control method was confirmed.
Yamano, Hidemasa; Futagami, Satoshi; Ando, Masanori
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 7 Pages, 2022/08
This study has conducted a detailed structural analysis of a reactor vessel (RV) in a loop-type sodium-cooled fast reactor using a general-purpose finite element analysis code, FINAS/STAR, to understand its deformation behavior under extremely high temperature conditions and to identify the areas which should be focused to mitigate impacts of failure. The RV was heated from the normal operation condition to the sodium boiling temperature in the upper sodium plenum during 20 hours assuming depressurization. The analysis has revealed less significant stress and strain which were sufficiently lower than failure criteria. The upper body of RV was identified as the important area in terms of mitigation of structural failure. The RV was eventually deformed downward about 16 cm, resulting in no failure. This effect contributes to maintaining RV sodium level in a long term, thereby enhancing the RV resilience.
Yamano, Hidemasa; Kubo, Shigenobu; Kan, Taro*; Shibata, Akihiro*; Hourcade, E.*; Dirat, J. F.*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 10 Pages, 2022/08
In this paper, the approach to event tree development and the scope of the event tree analysis were described with key points on core catcher loading. For the analytical conditions, two core catcher loading conditions were given as bounding and conservative cases. For important heading of the event tree, key important phenomena were included: strong back design, fuel-coolant interaction and quench in the sodium plenum design, jet attack, criticality and coolability on the core catcher. In this paper, preliminary trial quantification was attempted using a probability ranking table which is based on engineering judgement. This event tree analysis has identified the dominant sequence, and clarified the effect of the core catcher loading and effectiveness of design measures. This study suggests that the criticality measure is very important for the core catcher study.
Yamano, Hidemasa; Kubo, Shigenobu; Sasa, Kyohei*; Shibata, Akihiro*; Hourcade, E.*; Dirat, J. F.*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 9 Pages, 2022/08
This paper describes coolability evaluations of a debris bed with a variety of decay heat removal system (DHRS) operating conditions with a whole vessel model assuming fuel accumulation on the core catcher in a short term. The evaluation tool is a one-dimensional plant dynamics code, Super-COPD, with a debris bed module. The coolability evaluations have indicated that the current core catcher design secures sufficient natural circulation flows around the core catcher to ensure the debris bed cooling when at least one circuit of DHRS was activated. Sensitivity analyses under a pessimistic condition have shown that the debris bed is coolable with at least one circuit of improved DHRS even if most of fuel accumulates on the core catcher in a short term.
Kang, Z.; Okuda, Yukihiko; Nishida, Akemi; Tsubota, Haruji; Li, Y.
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 9 Pages, 2022/08
Most of the empirical formulas that have been proposed seeking to quantitatively investigate local damage to reinforced concrete (RC) structures caused by a rigid projectile impact. These formulas have been derived based on impact tests performed normal to the target structure, while only a few impact tests involving soft projectile to the target structure have been studied. The purpose of this study is to develop a local damage evaluation method that takes into account the oblique impact due to soft projectile, which should be considered in realistic impact conditions. In this paper, we compare the test results with the analytical results to examine and validate the parameter setting of analytical method for evaluating local damage in RC panel. The obtained knowledge is presented.
Kubo, Kotaro; Fujiwara, Keita*; Tanaka, Yoichi; Hakuta, Yuto*; Arake, Daisuke*; Uchiyama, Tomoaki*; Muramatsu, Ken*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 8 Pages, 2022/08
After the Fukushima Daiichi Nuclear Power Plant accident, the importance of conducting probabilistic risk assessments (PRAs) of external events, especially seismic activities and tsunamis, was recognized. The Japan Atomic Energy Agency has been developing a computational methodology for seismic PRA, called the direct quantification of fault tree using Monte Carlo simulation (DQFM). When appropriate correlation matrices are available for seismic responses and capacities of components, the DQFM makes it possible to consider the effect of correlated failures of components connected through AND and/or OR gates in fault trees, which is practically difficult when methods using analytical solutions or multidimensional numerical integrations are used to obtain minimal cut set probabilities. The usefulness of DQFM has already been demonstrated. Nevertheless, a reduction of the computational time of DQFM would allow the large number of analyses required in PRAs conducted by regulators and/or operators. We; therefore, performed scoping calculations using three different approaches, namely quasi-Monte Carlo sampling, importance sampling, and parallel computing, to improve calculation efficiency. Quasi-Monte Carlo sampling, importance sampling, and parallel computing were applied when calculating the conditional core damage probability of a simplified PRA model of a pressurized water reactor, using the DQFM method. The results indicated that the quasi-Monte Carlo sampling works well at assumed medium and high ground motion levels, importance sampling is suitable for assumed low ground motion level, and that parallel computing enables practical uncertainty and importance analysis. The combined implementation of these improvements in a PRA code is expected to provide a significant acceleration of computation and offers the prospect of practical use of DQFM in risk-informed decision-making.
Onoda, Yuichi; Uchita, Masato*; Tokizaki, Minako*; Okazaki, Hitoshi*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08
The safety analyses were carried out to confirm the sufficiency of the function of the plant protection system against the pump/diagrid link rupture. The target plant is a pool-type SFR of about 600 MWe class equipped with an axially homogeneous core currently under development in Japan. In the pool-type SFR, the primary system piping connects primary pump and the high-pressure sodium plenum located at the inlet of fuel sub-assemblies and called "pump/diagrid link". Because this piping is submerged in the reactor vessel, it is difficult to detect small scale sodium leakage in this piping, and thus a certain large pipe break like guillotine should be assumed and evaluated as a design basis event. In order to confirm the detectability of pump/diagrid link rupture by safety protection system signals, a series of analyses of the guillotine break for a pump/diagrid link were carried out. Sensitivity study had also been performed to consider the uncertainty of the reactivity coefficient in the analyses. The sufficiency of the function of the plant protection system against the pump/diagrid link rupture was confirmed by the analysis results that at least two signals are transmitted for the detection of the event, which is the development target of the plant protection system in pool-type SFR.
Choi, B.; Nishida, Akemi; Shiomi, Tadahiko; Kawata, Manabu; Li, Y.
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08
In the seismic evaluation of nuclear facility buildings, basemat uplift-the phenomenon during which the bottom of the basemat of a building partially rises from the ground owing to overturning moments during earthquakes-is a very important aspect because it affects not only structural strength and integrity, but also the response of equipment installed in the building. However, there are not enough analytical studies on the behavior of buildings with a low ground contact ratio due to basemat uplift during earthquakes. In this study, we conducted a simulation using a three-dimensional finite element model from past experiments on basemat uplift; further, we confirmed the validity of this approach. In order to confirm the difference in the analytical results depending on the analysis code, the simulation was performed under the same analytical conditions using the three analysis codes, which are E-FrontISTR, FINAS/STAR and TDAPIII, and the obtained analysis results were compared. Accordingly, we investigated the influence of the difference in adhesion on the structural response at low ground contact ratio. In addition, we confirmed the effects of significant analysis parameters on the structural response via sensitivity analysis. In this paper, we report the analytical results and insights obtained from these investigations.
Yada, Hiroki; Takaya, Shigeru; Morohoshi, Kyoichi*; Yokoi, Shinobu*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 9 Pages, 2022/08
In order to develop rationalized maintenance plans of nuclear power plants, it is necessary to consider characteristics of each plant. For sodium-cooled fast reactor (SFR) plants, there are constraints on inspections due to the specialty that sodium equipment needs to be handled, which is one of the important points when considering rationalization of maintenance. Therefore, we previously proposed a maintenance optimization scheme based on the System Based Code (SBC) concept. One of proposed scheme goals is to develop detailed procedures of preparing a rationalized maintenance plan. In this study, the procedures to determine inspections for potential degradation and additional inspections in terms of defense-in-depth have been further clarified. Furthermore, the modified maintenance optimization scheme is also illustrated by a quantitative trial evaluation of the core support structure of the next SFR under development in Japan.
Koie, Ryusuke*; Kawaguchi, Munemichi*; Miyahara, Shinya*; Uno, Masayoshi*; Seino, Hiroshi
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 4 Pages, 2022/08
In order to investigate removal mechanisms of cesium aerosol from noble gas bubble in sodium pool, we performed a water simulation test to measure the decontamination factors of simulant aerosols with nitrogen gas bubbles rising through the water pool. As a result, it was found that the decontamination factors increased with the increase in the aerosol diameter and the water pool depth.
Matsushita, Kentaro; Ezure, Toshiki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Tanaka, Masaaki
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 8 Pages, 2022/08
Development of evaluation method for cover gas entrainment (GE) by vortices generated at free surface in upper plenum of sodium-cooled fast reactor (SFR) is required. GE evaluation tool, named StreamViewer, based on method using numerical results of three-dimensional computational fluid dynamics analysis for loop-type SFRs has been developed. In this study, modification of evaluation method of StreamViewer to rationalize conservativeness in evaluation results was examined by identifying vortex center lines and calculating three-dimensional distribution of pressure decrease along vortex center lines. The applicability of modified method was checked using water experimental result in rectangular open channel where unsteady vortices are generated. As the result, it was indicated that evaluation results on gas core depth which were excessive in current method were improved in modified method, and it is confirmed that modified method may discriminate onset of GE with appropriate criteria.
Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 10 Pages, 2022/08
To confirm the applicability of the reactivity model, the authors have been conducting the benchmark exercises of the unprotected loss of heat sink event tests in a pool-type experimental fast reactor EBR-II. In the blind phase in the benchmark analyses using the plant dynamics analysis (1D) code in which the cold pool was modeled by means of the perfect mixing volume, it was found the increase of the core inlet temperature was evaluated lower than that of the measured data and the feedback reactivity was underestimated, because the thermal stratification in the cold pool was ignored. Then, the detailed model of the cold pool for the computational fluid dynamics (CFD) code was introduced and the 1D-CFD codes coupling method was applied to the benchmark analyses. It was confirmed that both the thermal stratification in the cold pool and the increase of the core inlet temperature were successfully reproduced.
Nakashima, Risako*; Koike, Akari*; Sakai, Takaaki*; Doda, Norihiro; Tanaka, Masaaki
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08
In development of a quantitative risk assessment method to the external hazards for a sodium-cooled fast reactor, a dynamic PRA using the Continuous Markov chain Monte Carlo (CMMC) method was performed to evaluate the effect of global warming on the probability of exceeding the temperature limit as a core damage factor. There is a possibility that the amount of snowfall in abnormal snowfall events will increase due to global warming in the future. A hazard curve of snowfall considering global warming was developed. The results show that the probability of exceeding the temperature limit is increased by the abnormal snowfall events due to global warming.
