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Maruyama, Shuhei; Yamamoto, Akio*; Endo, Tomohiro*
Annals of Nuclear Energy, 205, p.110591_1 - 110591_13, 2024/09
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)Rodriguez, D.; Rossi, F.; Takahashi, Tone
IEEE Transactions on Nuclear Science, 71(3), p.255 - 268, 2024/03
Times Cited Count:0 Percentile:0.00(Engineering, Electrical & Electronic)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.
Wakui, Takashi; Takagishi, Yoichi*; Futakawa, Masatoshi; Tanabe, Makoto*
Jikken Rikigaku, 23(2), p.168 - 174, 2023/06
Cavitation damage on the inner surface of the mercury target for the spallation neutron source occurs by proton bombarding in mercury. The prediction method of the cavitation damage using Monte Carlo simulations was suggested taking variability of the bubble core position and impact pressure distribution into account. The impact pressure distribution was estimated using the inverse analysis with Bayesian optimization was conducted with comparison between cavitation damage distribution obtained from experiment and the cumulative plastic strain distribution obtained from simulation. The average value and spread of maximum impact pressure estimated assuming the Gaussian distribution were 3.1 GPa and 1.2 m, respectively. Simulation results reproduced experimental results and it can be said that this evaluation method is useful.
Kubo, Kotaro; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Journal of Nuclear Science and Technology, 60(4), p.359 - 373, 2023/04
Times Cited Count:8 Percentile:83.23(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.
Rodriguez, D.; Koizumi, Mitsuo; Rossi, F.; Takahashi, Tone
Dai-43-Kai Nihon Kaku Busshitsu Kanri Gakkai Nenji Taikai Kaigi Rombunshu (Internet), 4 Pages, 2022/11
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.
Kubo, Kotaro; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Journal of Nuclear Science and Technology, 59(3), p.357 - 367, 2022/03
Times Cited Count:6 Percentile:56.19(Nuclear Science & Technology)Dynamic probabilistic risk assessment (PRA), which handles epistemic and aleatory uncertainties by coupling the thermal-hydraulics simulation and probabilistic sampling, enables a more realistic and detailed analysis than conventional PRA. However, enormous calculation costs are incurred by these improvements. One solution is to select an appropriate sampling method. In this paper, we applied the Monte Carlo, Latin hypercube, grid-point, and quasi-Monte Carlo sampling methods to the dynamic PRA of a station blackout sequence in a boiling water reactor and compared each method. The result indicated that quasi-Monte Carlo sampling method handles the uncertainties most effectively in the assumed scenario.
Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.308 - 315, 2020/10
Dynamic probabilistic risk assessment (PRA) is a method for improving the realism and completeness of conventional PRA. However, enormous calculation costs are incurred by these improvements. One solution is to select an appropriate sampling method. In this paper, we applied the Monte Carlo, Latin hypercube, grid-point, and quasi-Monte Carlo sampling methods to the dynamic PRA of a simplified accident sequence and compared the results for each method. Quasi-Monte Carlo sampling was found to be the most effective method in this case.
Hashimoto, Shintaro; Sato, Tatsuhiko
Journal of Nuclear Science and Technology, 56(4), p.345 - 354, 2019/04
Times Cited Count:7 Percentile:57.46(Nuclear Science & Technology)Particle transport simulations based on the Monte Carlo method have been applied to shielding calculations. Estimation of not only statistical uncertainty related to the number of trials but also systematic one induced by unclear physical quantities is required to confirm the reliability of calculated results. In this study, we applied a method based on analysis of variance to shielding calculations. We proposed random- and three-condition methods. The first one determines randomly the value of the unclear quantity, while the second one uses only three values: the default value, upper and lower limits. The systematic uncertainty can be estimated adequately by the random-condition method, though it needs the large computational cost. The three-condition method can provide almost the same estimate as the random-condition method when the effect of the variation is monotonic. We found criterion to confirm convergence of the systematic uncertainty as the number of trials increases.
Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa
JAEA-Data/Code 2016-018, 421 Pages, 2017/03
In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants, etc. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions.
Maeda, Makoto; Komeda, Masao; Tobita, Hiroshi; Ozu, Akira; Kureta, Masatoshi; Bogucarska, T.*; Crochemore, J. M.*; Varasano, G.*; Pedersen, B.*
Dai-37-Kai Kaku Busshitsu Kanri Gakkai Nihon Shibu Nenji Taikai Rombunshu (CD-ROM), 7 Pages, 2017/02
JAEA and EC/JRC are carrying out collaborative research to develop NDA techniques that can be utilized for quantification of high radioactive special nuclear materials such as spent fuel and next generation minor actinide fuels. In the research, reliability of neutron transport codes is important because it is utilized for design and development of a demonstration system of next-generation Differential Die-away (DDA) technique in JAEA. In order to evaluate the reliability, actual neutron flux distribution in a sample cavity was examined in PUNITA device using JRC type DDA technique and JAWAS-T device using JAEA type DDA technique, and then the measurement results were compared with the simulation results obtained by the neutron transport codes. The neutron flux distribution in the target matrix was also examined in the PUNITA and compared with the simulation results. We report on the measurement and simulation results of the neutron flux distribution and evaluation results of the reliability of the neutron transport codes.
Maeda, Makoto; Komeda, Masao; Tobita, Hiroshi; Ozu, Akira; Kureta, Masatoshi; Bogucarska, T.*; Crochemore, J. M.*; Varasano, G.*; Pedersen, B.*
Proceedings of INMM 57th Annual Meeting (Internet), 9 Pages, 2016/07
The JAEA and EC/JRC have started collaborative research to develop a technique that can be utilized for quantification of high radioactive special nuclear materials such as next generation minor actinide fuels. In the study of a Differential Die-Away (DDA) technique, which is one of the techniques to be improved in the collaborative research, JRC type and JAEA type DDA techniques are compared. In the JRC type DDA technique, large amount of thermal neutron is generated using D-T neutron generator and graphite moderator to accomplish high detection sensitivity for small amount of fissile material. On the other hand, in JAEA type, relatively hard neutron spectrum and moderation of neutron in the target matrix are utilized to minimize position dependence of detection efficiency. Estimation of the neutron field is important to evaluate the performance of the system in DDA technique. The purpose of this study is to validate simulation results by experimental results and evaluate neutron flux distribution in the system by the simulation and the experiment. In this paper, we present the evaluation results of the neutron flux distributions in PUNITA which utilizes JRC type DDA technique and JAWAS-T which utilizes JAEA type DDA technique obtained by Monte Carlo simulation and activation method.
Fukushima, Masahiro; Kitamura, Yasunori; Kugo, Teruhiko; Okajima, Shigeaki
Journal of Nuclear Science and Technology, 53(3), p.406 - 424, 2016/03
Times Cited Count:12 Percentile:69.28(Nuclear Science & Technology)Tsutsumi, Masahiro; Tanimura, Yoshihiko
Nuclear Instruments and Methods in Physics Research A, 557(2), p.554 - 560, 2006/02
Times Cited Count:16 Percentile:72.08(Instruments & Instrumentation)A new cerium-doped LaCl(Ce) scintillator is evaluated with respect to the application in environmental -ray dosimetry and spectrometry. The scintillator is very attractive for -ray spectrometry in the case of high count rate, because it has excellent energy resolution and fast decay time. The performance characteristics of a scintillator with a 25.4 mm 25.4 mm LaCl(Ce) crystal are studied and compared to those of a NaI(Tl) scintillator with the same size crystal. Acquired pulse-height spectra are converted to dose rates by using the G(E) function method. Though the LaCl(Ce) crystal itself produces a rather high background in the crystal itself, the scintillator provides good energy information and dose-rate readings from low to high-level (several mGy/h) by subtracting the self-background. The properties of LaCl(Ce) scintillator suggest that the scintillator could be a promising candidate for monitoring at high-dose levels as in emergencies, as well as at ordinary levels of background radiation.
Asano, Yoshihiro; Sugita, Takeshi*; Suzaki,Takenori; Hirose, Hideyuki
Radiation Protection Dosimetry, 116(1-4), p.284 - 289, 2005/12
Times Cited Count:0 Percentile:0.00(Environmental Sciences)no abstracts in English
Kugo, Teruhiko; Mori, Takamasa
Proceedings of International Topical Meeting on Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications (M&C 2005) (CD-ROM), 10 Pages, 2005/09
A new deterministic transport code based on the method of characteristics (MOC) has been developed for heterogeneous transport calculations in core design of innovative reactors which have complex structures. We have investigated the capability of the MOC code for general geometry with an unstructured geometry PWR problem. The comparison of the results with accurate Monte Carlo calculation results by GMVP has confirmed that the MOC code produces satisfactory results and has a capability to treat unstructured geometry.
Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki
JAERI 1348, 388 Pages, 2005/06
To realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector supercomputers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them.
Nagaya, Yasunobu; Mori, Takamasa
Journal of Nuclear Science and Technology, 42(5), p.428 - 441, 2005/05
Times Cited Count:62 Percentile:95.96(Nuclear Science & Technology)A new method to estimate a change in the effective multiplication factor due to the perturbed fission source distribution has been proposed for Monte Carlo perturbation calculations with the correlated sampling and differential operator sampling techniques. The method has been implemented into the MVP code for verification. Simple benchmark problems have been set up for fast and thermal systems and the applicability of the method has been verified with the problems. In consequence, it has been confirmed that the method is very effective to estimate the change. It has been also shown that there are some cases where the perturbed source effect is significant and the change in reactivity cannot be estimated accurately without taking the effect into account. Even in such cases, the new method can estimate the perturbed source effect and the estimation of the change in reactivity has been remarkably improved.
Matsubayashi, Masahito; Hibiki, Takashi*; Mishima, Kaichiro*; Yoshii, Koji*; Okamoto, Koji*
Nuclear Instruments and Methods in Physics Research A, 533(3), p.481 - 490, 2004/11
Times Cited Count:5 Percentile:36.07(Instruments & Instrumentation)The validity of a fast neutron radiography quantification method, the -scaling method, which was originally proposed for thermal neutron radiography was examined with Monte Carlo calculations and experiments conducted at the YAYOI fast neutron source reactor. Water and copper were selected as comparative samples for a thermal neutron radiography case and a dense object, respectively. Although different characteristics on effective macroscopic cross-sections were implied by the simulation, the -scaled experimental results with the fission neutron spectrum cross-sections were well fitted to the measurements for both the water and copper samples. This indicates that the -scaling method could be successfully adopted for quantitative measurements in fast neutron radiography.