Faculty Profiles - YAMAMOTO, Akio

Papers - YAMAMOTO, Akio

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Loading Pattern Optimization for LWRs using Monte Carlo Tree Search

Kasama R., Yamamoto A., Endo T.

Proceedings of the International Conference on Physics of Reactors Physor 2024 page： 2490 - 2499 2024

Language：English Publishing type：Research paper (scientific journal) Publisher：Proceedings of the International Conference on Physics of Reactors Physor 2024

Inspired by recent breakthroughs in AI for games, we transform the fuel loading pattern optimization problem into a game tree search problem and adopt Monte Carlo Tree Search (MCTS) to solve it. The performance of MCTS is compared with the conventional optimization methods, such as the random search and simulated annealing (SA), for the loading pattern optimization of a typical pressurized water reactor. The results demonstrate the superior efficiency of MCTS over the conventional methods. MCTS consistently produces high-quality patterns and effectively avoids the common problem of falling into local minima. By using techniques rooted in AI, we are opening up a new way of solving complex problems in nuclear engineering, offering the prospect of more reliable and efficient core designs. This innovative approach will not only benefit the nuclear field, but also serve as a bridge between optimization and artificial intelligence.

DOI： 10.13182/PHYSOR24-43466

Scopus

Investigation of Subcriticality Monitoring Method using Improved Simplest Reactivity Estimator with Bilateral Filter

Moribe T., Endo T., Yamamoto A., Kaneko J.H.

Proceedings of the International Conference on Physics of Reactors Physor 2024 page： 97 - 106 2024

Language：English Publishing type：Research paper (scientific journal) Publisher：Proceedings of the International Conference on Physics of Reactors Physor 2024

In the retrieval of fuel debris from the Fukushima Daiichi nuclear power plant, there are the following issues: lack of information needed to inversely estimate the subcriticality; and the limitation of neutron detector size. In order to address these issues, we propose a reactivity estimation method that combines the improved simplest reactivity estimator (SRE) and the least squares inverse kinetics method. This study aims to estimate the reactivity in dollar units in a data-driven manner only from the measured neutron count rate without point kinetics parameters. Furthermore, the usefulness of the bilateral filter in combination with the median for the count rate is also investigated to mitigate statistical errors in the reactivity estimation results due to the statistical fluctuation and outlier in the measured count rate data. To verify our proposed method, point kinetics calculation is performed to simulate the time series data of the neutron count rate during a stepwise transient from a deep subcritical state to the shallower state. Here, the magnitude of the count rate at the initial state is parametrically varied to investigate the effect of the statistical count rate fluctuation on the estimated reactivity. As a result, it is observed that the estimated reactivity tends to be overestimated (i.e., closer to critical) as lower count rate data are used in our proposed method. Thus, our proposed method is considered practical from the perspective of criticality control. Moreover, it is clarified that the combinational use of the bilateral filter and the median can effectively reduce the error of the estimated reactivity due to the statistical of the count rate. Consequently, this feasibility study demonstrates our proposed method for monitoring the fluctuation reactivity change even under the low neutron count rate condition.

DOI： 10.13182/PHYSOR24-43487

Scopus

Statistical Error Estimation of Autocorrelation Method using Circular Block Bootstrap Method

Hirota R., Endo T., Yamamoto A., Watanabe K., Kaneko J.H.

Proceedings of the International Conference on Physics of Reactors Physor 2024 page： 78 - 86 2024

Language：English Publishing type：Research paper (scientific journal) Publisher：Proceedings of the International Conference on Physics of Reactors Physor 2024

The autocorrelation method is a subcriticality measurement technique based on the reactor noise analysis method. In this method, the prompt neutron decay constant can be obtained from the exponential decay of the autocorrelation of successively detected neutron counts in a target subcritical system. To simply estimate the statistical error of the prompt neutron decay several measurements of the reactor noise are required, although the total measurement time is inevitably long constant, for typical systems where the reactor noise analysis method is applied. Therefore, the purpose of this study is to investigate the applicability of the circular block bootstrap method in order to estimate the statistical error of the prompt neutron decay constant obtained by the autocorrelation method using a single reactor noise measurement. The bootstrap-based statistical error estimation method is validated using the time series data of reactor noise measurements at UTR-KINKI in a shutdown state with the inherent neutron source in the uranium-aluminum fuel. Consequently, this study demonstrates that the circular block bootstrap method can be also utilized for the autocorrelation method, to estimate the confidence interval of the prompt neutron decay constant as the statistical error. Namely, a single reactor noise measurement can be effectively reused for error estimation instead of multiple measurements.

DOI： 10.13182/PHYSOR24-43468

Scopus

Verification of Neutronics Analysis Method using CBZ and GENESIS for a Prismatic High-Temperature Gas-Cooled Reactor

Yamamoto A., Chiba G.

Proceedings of the International Conference on Physics of Reactors Physor 2024 page： 1572 - 1580 2024

Language：English Publishing type：Research paper (scientific journal) Publisher：Proceedings of the International Conference on Physics of Reactors Physor 2024

A neutronics calculation method for prismatic-type high-temperature gas-cooled reactors is developed using a general neutronics analysis code CBZ and a 2D/3D multi-group neutron transport code GENESIS.Fuel element and a simplified two-dimensional core of the high-temperature engineering test reactor (HTTR) is considered as the benchmark problems and the calculation results of CBZ and GENESIS are compared with those obtained by the continuous energy Monte Carlo code MVP.K-effective, fuel element/fuel compact-wise power distributions are compared.The results of the comparison indicate that CBZ and GENESIS accurately reproduce the reference results obtained by MVP.

DOI： 10.13182/PHYSOR24-43456

Scopus

THERMAL CUT-OFF ENERGY FOR ACCURATE ANALYSIS OF PRISMATIC HIGH-TEMPERATURE GAS-COOLED REACTOR

Takeda S., Kitada T., Yamamoto A., Yamaji K., Koike H., Asano K.

Proceedings of 2024 31st International Conference on Nuclear Engineering ICONE 2024 Vol. 2 2024

Language：English Publishing type：Research paper (scientific journal) Publisher：Proceedings of 2024 31st International Conference on Nuclear Engineering ICONE 2024

In this study, the appropriate thermal cut-off energy for neutron transport calculations in High-Temperature Gas-cooled Reactor (HTGR) is investigated. The research employs the Monte Carlo calculation code MVP3.0 to evaluate the impact of changing thermal cut-off energy on the multiplication factor in the single fuel element geometries, focusing on variations in fuel and moderator temperatures at both the Beginning Of Life (BOL) and End Of Life (EOL). Up-scattering effect mainly caused by graphite used in HTGR as moderator has strong relation to the thermal cut-off energy. The findings show that when the thermal cut-off energy is set at 20 eV, the multiplication factor tends to be underestimated, especially at high fuel temperature. Furthermore, the study confirms that when the thermal cut-off energy is at or above 30 eV, the variations in multiplication factors remain within approximately ±50 pcm.

DOI： 10.1115/ICONE31-130801

Scopus

Study on Fusion Safety Concept Based on Nuclear Safety (2) Defense in Depth in Fusion Safety

Yamahista Y., Endo T., Yamamoto A., Ino T.

Transactions of the American Nuclear Society Vol. 131 ( 1 ) page： 299 - 302 2024

Study on Fusion Safety Concept Based on Nuclear Safety (1) Fusion Safety Objective and Principles

Yamahista Y., Endo T., Yamamoto A., Ino T.

Transactions of the American Nuclear Society Vol. 131 ( 1 ) page： 292 - 295 2024

Loading Pattern Optimization with Deep Learning and Monte-Carlo Tree Search

Kasama R., Yamamoto A., Endo T.

Transactions of the American Nuclear Society Vol. 130 ( 1 ) page： 1044 - 1047 2024

Data Assimilation Using Deterministic Sampling Method to Selectively Reduce Uncertainty due to Thermal Neutron Scattering Law for Light Water

Harada Y., Yamaguchi H., Endo T., Yamamoto A., Tada K.

Transactions of the American Nuclear Society Vol. 130 ( 1 ) page： 758 - 762 2024

A Simple Resonance Calculation Approach for Space-dependent Self-shielding

Yamamoto A., Endo T.

Transactions of the American Nuclear Society Vol. 131 ( 1 ) page： 1032 - 1035 2024

2D Cylindrical SP3 Transport Calculation Using Finite Volume Method with Cross-Diffusion Treatment for Unstructured Meshes

Endo T., Yamamoto A.

Transactions of the American Nuclear Society Vol. 131 ( 1 ) page： 1053 - 1056 2024

THERMAL CUT-OFF ENERGY FOR ACCURATE ANALYSIS OF PRISMATIC HIGH-TEMPERATURE GAS-COOLED REACTOR

Takeda, S; Kitada, T; Yamamoto, A; Yamaji, K; Koike, H; Asano, K

PROCEEDINGS OF 2024 31ST INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, VOL 2, ICONE31 2024 2024

Sensitivity analysis of risk assessment for continuous Markov process Monte Carlo method using correlated sampling method

Morishita, Y; Yamamoto, A; Endo, T

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY Vol. 60 ( 12 ) page： 1573 - 1585 2023.12

Language：English Publishing type：Research paper (scientific journal) Publisher：Journal of Nuclear Science and Technology

The correlated sampling method is applied to the continuous Markov-chain Monte-Carlo (CMMC) method to efficiently perform sensitivity analysis of input parameters such as the failure rate of safety components. In the correlated sampling method, the original and the perturbed samples are assumed to trace an identical accident sequence, but the weight of the perturbed sample is adjusted to incorporate the variation of input data. The present method is applied to the sensitivity analysis of the safety evaluation of spent fuel pools. The result indicates that the sensitivity analysis for the CMMC coupling method can be efficiently carried out using the correlated sampling method.

DOI： 10.1080/00223131.2023.2231464

Web of Science

Scopus

Impact of nuclear data revised from JENDL-4.0 to JENDL-5 on PWR spent fuel nuclide composition Open Access

Watanabe, T; Tada, K; Endo, T; Yamamoto, A

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY Vol. 60 ( 11 ) page： 1386 - 1396 2023.11

Language：English Publishing type：Research paper (scientific journal) Publisher：Journal of Nuclear Science and Technology

The burnup calculations for estimating the nuclide composition of the spent fuel are highly dependent on nuclear data. Many nuclides in the latest version of the Japanese Evaluated Nuclear Data Library JENDL-5 were modified from JENDL-4.0 and the modification affects the burnup calculations. This study confirmed the validity of JENDL-5 in the burnup calculations. The PIE data of Takahama-3 was used for the validation. The effect of modifications of the parameters, e.g. cross sections and fission yields, from JENDL-4.0 to JENDL-5 on the nuclide compositions was quantitatively investigated. The calculation results showed that JENDL-5 has a similar performance to JENDL-4.0. The calculation results also revealed that the modifications of the cross sections of actinide nuclides, fission yields, and thermal scattering low data of hydrogen in H<inf>2</inf>O affected the nuclide compositions of PWR spent fuels.

DOI： 10.1080/00223131.2023.2201603

Web of Science

Scopus

An estimation method for an unknown covariance in cross-section adjustment based on unbiased and consistent estimator

Maruyama, S; Endo, T; Yamamoto, A

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY Vol. 60 ( 11 ) page： 1372 - 1385 2023.11

Language：English Publishing type：Research paper (scientific journal) Publisher：Journal of Nuclear Science and Technology

A new estimation method of an unknown covariance in the cross-section adjustment method for the development of an application library is proposed. The unknown covariance is defined by the difference between the true covariance (the population covariance) and a prior covariance assumed by an analyst. The unknown covariance is estimated using an empirical covariance consistent with the observed data. To estimate the unknown covariance, an unbiased and consistent estimator in regression analysis has been incorporated into the conventional cross-section adjustment. This estimator does not require assumptions for the probability distribution of the observation data. The statistical properties of this estimator were numerically verified. In addition, the effectiveness of the proposed method was confirmed by another numerical test using actual integral experimental data. In the second numerical test, the modeling uncertainty (covariance) due to the deterministic analysis method was assumed to be unknown. The results show that the proposed method can practically estimate the unknown covariance and adjusted cross-sections using only prior information on covariances.

DOI： 10.1080/00223131.2023.2203707

Web of Science

Scopus

Development of ACE file perturbation tool using FRENDY

Tada, K; Kondo, R; Endo, T; Yamamoto, A

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY Vol. 60 ( 6 ) page： 624 - 631 2023.6

Language：English Publishing type：Research paper (scientific journal) Publisher：Journal of Nuclear Science and Technology

The sensitivity analysis and the uncertainty quantification play an important role in improving the evaluated nuclear data library. The current computational performance enables us to perform sensitivity analysis and uncertainty quantification using continuous energy Monte Carlo calculation codes. The ACE file perturbation tool was developed for these calculations using modules of FRENDY. This tool perturbs the microscopic cross section, the number of neutrons per fission, and the fission spectrum. The uncertainty quantification using the random sampling method is also available if the user prepares the covariance matrices. The uncertainty of k<inf>eff</inf> using the ACE file perturbation tool was compared to the current sensitivity analysis codes SCALE/TSUNAMI and MCNP/KSEN. The comparison results indicated that the random sampling method using this tool accurately estimates the uncertainty of k<inf>eff</inf>. The ACE file perturbation tool is included in the current version of FRENDY.

DOI： 10.1080/00223131.2022.2130463

Web of Science

Scopus

Comparison of internal boundary conditions for optical diffusion calculations considering reflection and refraction Open Access

Amano T., Endo T., Yamamoto A.

Optics Continuum Vol. 2 ( 7 ) page： 1540 - 1560 2023.6

Language：English Publishing type：Research paper (scientific journal) Publisher：Optics Continuum

A method to treat the internal boundary condition in an optical diffusion calculation is proposed and is compared with the conventional methods. One of the existing internal boundary conditions is Haskel’s method, which uses the effective reflection coefficient for partial currents. However, Haskel’s method ignores incoming partial currents from the adjacent mesh in its derivation. As a result, the accuracy at the internal boundary is lower. This paper proposes a method to improve the accuracy by iteratively updating the effective reflection coefficient for partial current. The proposed method was applied to the benchmark calculations on a one-dimensional slab geometry and its accuracy was confirmed by comparing it with the reference solution obtained by the Monte Carlo code MCML, along with the previously proposed Haskel’s method and Aronson’s method. As a result, it was confirmed that the proposed method is more accurate than Haskel’s method at the internal boundary and gives the same result as Aronson’s method. The convergence of the effective reflection coefficient using iterative calculations in the proposed method was good.

DOI： 10.1364/OPTCON.492445

Open Access

Scopus

Efficient reduced order model based on the proper orthogonal decomposition for time-dependent MOC calculations Open Access

Tsujita, K; Endo, T; Yamamoto, A

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY Vol. 60 ( 3 ) page： 343 - 357 2023.3

Language：English Publishing type：Research paper (scientific journal) Publisher：Journal of Nuclear Science and Technology

An efficient reduced order model (ROM) for time-dependent transport calculations using the method of characteristics (MOC) is proposed. In the present ROM, the flux distributions and the net neutron currents between the adjacent unstructured mesh regions are taken from the MOC solution. Then, the coefficient matrices for the MOC-equivalent diffusion equation are reconstructed from them. The proper orthogonal decomposition (POD) is applied for the MOC-equivalent diffusion equation to reduce the degree of freedoms (DOFs) using the orthogonal bases obtained by the singular value decomposition (SVD) for the sampled MOC solution. The accuracy and computation time of the present ROM are verified in the C5G7-TD 2D benchmark problem. The calculation results show that the present ROM enables us approximately 5000–6000 times faster computation than the full order model (FOM) for kinetic calculations itself in the present calculation condition. The present method can be substituted as real-time simulations without the spatial homogenization when typical flux distributions and the net neutron currents of a target problem can be precalculated.

DOI： 10.1080/00223131.2022.2097963

Web of Science

Scopus

Nuclear data adjustment using a deterministic sampling method with unscented transformation

Fukui, Y; Endo, T; Yamamoto, A

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY Vol. 60 ( 3 ) page： 238 - 250 2023.3

Language：English Publishing type：Research paper (scientific journal) Publisher：Journal of Nuclear Science and Technology

A nuclear data adjustment method using a deterministic sampling method based on an unscented transform was developed, and its validity was confirmed through a twin experiment using a critical benchmark problem. Conventional nuclear data adjustment methods require sensitivity analysis using generalized perturbation theory or many forward calculations using stochastically sampled nuclear data. To address this issue, this study focused on unscented transform-based sampling (UTS), which is used in uncertainty quantification. Based on the UTS, perturbed nuclear data can be deterministically sampled to reproduce the population covariance matrix with minimum sample size. Therefore, UTS can significantly reduce the computational cost compared to conventional nuclear data adjustment using random sampling (RS). Furthermore, the UTS was improved to prevent the sampling of negative nuclear data while accurately reproducing the population covariance matrix. The proposed method was applied to the numerical experiment of Godiva, and the adjusted nuclear data were compared with those obtained using conventional methods. Consequently, it was demonstrated that UTS can adjust nuclear data at a lower computational cost than RS.

DOI： 10.1080/00223131.2022.2095051

Web of Science

Scopus

Theoretical Derivation of a Unique Combination Number Hidden in the Higher-Order Neutron Correlation Factors Using the Pal-Bell Equation Open Access

Endo, T; Nishioka, F; Yamamoto, A; Watanabe, K; Pyeon, CH

NUCLEAR SCIENCE AND ENGINEERING Vol. 197 ( 2 ) page： 176 - 188 2023.2

Language：English Publishing type：Research paper (scientific journal) Publisher：Nuclear Science and Engineering

The Pál-Bell equation is a backward-type master equation that describes the probability generating function (PGF) of neutron counts in a neutron multiplication system. Thanks to the Pál-Bell equation with the two-forked and the fundamental mode approximations, an analytical solution of PGF of neutron counts in a source-driven subcritical system can be theoretically derived. This theoretical derivation clarifies that the unique combination number of double factorial (2n−3)!! does exist in the ratio of the higher-order neutron correlation factors measured in a critical state even for any kind of fissile and moderator materials. Additionally, the unique combination numbers are experimentally validated for the order 3 ≤ n ≤ 6 through reactor noise measurements in actual subcritical systems. This knowledge can be used to judge whether a target system is in a deep subcritical state or to roughly estimate the magnitude of subcriticality, based on the factorial moments of the measured reactor noise in a zero-power state.

DOI： 10.1080/00295639.2022.2049992

Web of Science

Scopus

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Papers - YAMAMOTO, Akio