Assessment of probability of aircraft crashes for Nuclear Science Research Institute

Kamikawa, Yutaka; Suzuki, Makoto; Agake, Toshiki; Murakami, Takahiko; Morita, Yusuke; Shiina, Hidenori; Fukushima, Manabu; Hirane, Nobuhiko; Ouchi, Yasuhiro

JAEA-Technology 2023-030, 57 Pages, 2024/03

JAEA-Technology-2023-030.pdf:1.93MB

Owing to the publication of the latest data about aircraft crashes by Nuclear Regulation Authority (NRA), it was necessary to re-evaluate the probabilities of aircraft crashes for Nuclear Science Research Institute (NSRI). By using of the assessment method provided in "Regulatory Guide of the Assessment Standard for Probability of Airplane Crash on a Nuclear Power Reactor Facility", we re-evaluated the probabilities of aircraft crashes against the nuclear facilities in NSRI. As a result of the evaluations, the sum of the probabilities of aircraft crashes against Waste Treatment Facilities (maximum probability among all nuclear facilities in NSRI) is 5.6810 (times/(reactor year)) which is lower than 10 (times/(reactor year)) that is the assessment criterion whether aircraft crashes is considered to be "anticipated external human induced events" in design basis or not.

Atomic position and the chemical state of an active Sn dopant for Sn-doped -GaO(001)

Tsai, Y. H.*; Kobata, Masaaki; Fukuda, Tatsuo; Tanida, Hajime; Kobayashi, Toru; Yamashita, Yoshiyuki*

Applied Physics Letters, 124(11), p.112105_1 - 112105_5, 2024/03

https://doi.org/10.1063/5.0198160

Hierarchical deformation heterogeneity during Lders band propagation in an Fe-5Mn-0.1C medium Mn steel clarified through scanning electron microscopy

Koyama, Motomichi*; Yamashita, Takayuki*; Morooka, Satoshi; Yang, Z.*; Varanasi, R. S.*; Hojo, Tomohiko*; Kawasaki, Takuro; Harjo, S.

Tetsu To Hagane, 110(3), p.205 - 216, 2024/02

https://doi.org/10.2355/tetsutohagane.TETSU-2023-052

Estimation of continuous distribution of iterated fission probability using an artificial neural network with Monte Carlo-based training data

Tuya, D.; Nagaya, Yasunobu

Journal of Nuclear Engineering (Internet), 4(4), p.691 - 710, 2023/11

https://doi.org/10.3390/jne4040043

The Monte Carlo method is used to accurately estimate various quantities such as k-eigenvalue and integral neutron flux. However, when a distribution of a quantity is desired, the Monte Carlo method does not typically provide continuous distribution. Recently, the functional expansion tally and kernel density estimation methods have been developed to provide continuous distribution. In this paper, we propose a method to estimate a continuous distribution of a quantity using artificial neural network (ANN) model with Monte Carlo-based training data. As a proof of concept, a continuous distribution of iterated fission probability (IFP) is estimated by ANN models in two systems. The IFP distributions by the ANN models were compared with the Monte Carlo-based data and the adjoint angular neutron fluxes by the PARTISN code. The comparisons showed varying degrees of agreement or discrepancy; however, it was observed that the ANN models learned the general trend of the IFP distributions.

Simulation-based dynamic probabilistic risk assessment of an internal flooding-initiated accident in nuclear power plant using THALES2 and RAPID

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

https://doi.org/10.1177/1748006X221091604

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%.

Cavitation damage prediction in mercury target for pulsed spallation neutron source using Monte Carlo simulation

Wakui, Takashi; Takagishi, Yoichi*; Futakawa, Masatoshi

Materials, 16(17), p.5830_1 - 5830_16, 2023/09

https://doi.org/10.3390/ma16175830

Cavitation damage on the mercury target vessel is induced by proton beam injection in mercury. The prediction method of the cavitation damage using Monte Carlo simulations was proposed taking into account of the uncertainties of the position of cavitation bubbles and impact pressure distributions. The distribution of impact pressure attributed to individual cavitation bubble collapsing was assumed to be the Gaussian distribution, and the probability distribution of the maximum value of impact pressures was assumed to be three kinds of distributions; the delta function, the Gaussian and Weibull distributions. Two parameters were estimated using Bayesian optimization by comparing the distribution of the cavitation damage obtained from experiment with that of accumulated plastic strain obtained from the simulation. It was found that the results obtained using the Weibull distribution reproduced the actual cavitation erosion phenomenon better than the other results.

Behavior of FeCrAl-ODS cladding tube under loss-of-coolant accident conditions

Narukawa, Takafumi; Kondo, Keietsu; Fujimura, Yuki; Kakiuchi, Kazuo; Udagawa, Yutaka; Nemoto, Yoshiyuki

Journal of Nuclear Materials, 582, p.154467_1 - 154467_12, 2023/08

https://doi.org/10.1016/j.jnucmat.2023.154467

Application of quasi-Monte Carlo and importance sampling to Monte Carlo-based fault tree quantification for seismic probabilistic risk assessment of nuclear power plants

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

https://doi.org/10.1299/mej.23-00051

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.

Cavitation damage prediction in mercury target for pulsed spallation neutron sources by Monte Carlo simulation

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.

Introduce of friction model into fuel pin bundle deformation analysis code "BAMBOO"

Uwaba, Tomoyuki; Ito, Masahiro*; Ishitani, Ikuo*

JAEA-Technology 2023-006, 36 Pages, 2023/05

JAEA-Technology-2023-006.pdf:3.45MB

The BAMBOO code developed by the Japan Atomic Energy Agency is a computer code to analyze fuel pin bundle deformation in a fast reactor wire-spaced type fuel pin bundle subassembly. In this study we developed an analysis model to consider friction at the contact points between adjacent fuel pins, and at these between outermost fuel pins and a duct that are due to bundle-duct interaction. This model deals with friction forces at contact points in the contact and separation analysis of the code, and employs a convergent calculation where contact forces are gradually determined to avoid numerical instability when the friction occurs. Analyses of BAMBOO with the model showed very slight effects on the onset of contact between outer most pins and a duct, and on directions of pin displacements, within the range of practical friction coefficients.