Faculty Profiles - YAMAMOTO, Akio

Papers - YAMAMOTO, Akio

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Real-Time 3D Fine Spatial Mesh Kinetics Simulator using POD for Coupled Core

Ito K., Tsujita K., Endo T., Yamamoto A.

Proceedings of the International Conference on Physics of Reactors Physor 2024 page： 1622 - 1631 2024

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

Toward the development of the digital triplets to effectively educate the reactor physics experiment, the present study develops a real-time 3D fine spatial mesh kinetics simulator based on the proper orthogonal decomposition (POD).Our developed simulator, called Ikaros3D, was specifically designed for a coupled core to promote students' better understanding of the complicated behavior of the spatial variation of the dynamic reactivity depending on the neutron detector position.Through a verification test, we confirmed that the reduced order model (ROM) using POD was able to accurately and quickly simulate the variation in the total core power due to the operation of two different control rods, thanks to the pre-tabulated compressed coefficient matrices for the POD-based kinetics calculation.

DOI： 10.13182/PHYSOR24-43834

Scopus

Preliminary Study on Two-Dimensional SP3 Calculation Based on POD-Local/Global Iterations

Ito M., Yamamoto A., Endo T., Takeishi T., Kodama Y., Nagano H.

Proceedings of the International Conference on Physics of Reactors Physor 2024 page： 1612 - 1621 2024

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

As a preliminary study for an efficient heterogeneous neutron transport calculation, this study newly proposes a reduced order core calculation method by applying proper orthogonal decomposition (POD) to the SP3 calculation with the global/local iterations.The fine mesh SP3 calculation (local calculation) for each single assembly can be efficiently solved using POD, thanks to the matrix form-based SP3 equation and the relatively small scale of the simultaneous equations.By reconstructing the fine mesh distribution of neutron flux and partial neutron currents from the POD bases and the expansion coefficients estimated by the local calculation, the p-CMFD calculation is applied to the coarse mesh calculation (global calculation).These local and global calculations are iterated until convergence.Through the two-dimensional small and large UO2-MOX core calculations, our proposed method is verified by comparing with the reference result by the conventional fine mesh SP3 calculation.

DOI： 10.13182/PHYSOR24-43629

Scopus

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

Reconstruction of In-core Power Distribution Based on POD Using Ex-core Detector Signals

Urase Y., Endo T., Yamamoto A.

Proceedings of the International Conference on Physics of Reactors Physor 2024 page： 1642 - 1651 2024

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

A reconstruction method of in-core power distribution using the proper orthogonal decomposition (POD) and the signal from ex-core detectors is developed.In the present method, the neutron flux distribution is expanded in a small number of POD bases and expansion coefficients.Neutron transport calculations were performed with several calculation conditions to perturb the neutron flux distributions in the reactor core and the POD bases were obtained using the multiple neutron flux distributions as the snapshot data.The neutron counts of ex-core detectors were evaluated from in-core neutron flux distributions and the detector response coefficients.Considering the neutron the counts of ex-core detectors as measured signals, the expansion coefficients of POD that represent the in-core neutron flux distributions were determined.Assuming a simple two-dimensional core with nine regions, the reconstruction accuracy of in-core neutron flux distribution was verified for several conditions.The results indicate that the in-core neutron flux distributions can be reconstructed by the ex-core detector signals when the distance between the detectors and the core is appropriately chosen.

DOI： 10.13182/PHYSOR24-43478

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

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

Simplified Treatment of Coating Layers in TRISO Fuel in Statistical Geometry Method in Monte Carlo Calculation

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

Proceedings of the International Conference on Physics of Reactors Physor 2024 page： 1036 - 1047 2024

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

An improved sampling method for flight distance is proposed for Monte Carlo analysis of TRISO fuel particles using the statistical geometry method.The statistically uniform distribution of fuel particles, which is usually assumed as a default sampling method of flight distance of a neutron in graphite matrix, shows considerable bias on neutronics property when coating layers of a TRISO fuel particle are homogenized with graphite matrix.The proposed new sampling method almost resolves the difference between the homogenized coating model and the reference model that explicitly considers coating layers.By adopting the present method, the homogenized coating model can be used without significant loss of accuracy in the statistical geometry method.Computation time for a typical HTTR fuel compact cell with a continuous energy Monte Carlo code is reduced to 1/7 when the homogenized coating model is used.

DOI： 10.13182/PHYSOR24-43318

Scopus

Resonance Treatment of Double-Heterogeneous Fuel using a Deterministic Statistical Geometry Method in Heterogeneous Transport Calculation Code GALAXY-Z

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

Proceedings of the International Conference on Physics of Reactors Physor 2024 page： 1025 - 1035 2024

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

A resonance calculation using a deterministic statistical geometry (DSTG) method has been developed to efficiently treat double heterogeneity of coated fuel particles loaded in a high temperature gas-cooled reactor (HTGR). By the DSTG method, neutron flux spatial distribution can be calculated in a heterogeneous geometry containing randomly dispersed fuel particles and surrounding graphite matrix, which is consistent with STG in a Monte Carlo method. We applied the DSTG method to resonance calculations based on the equivalence theory and the ultra-fine-group energy treatment in the heterogeneous transport calculation code GALAXY-Z developed by Mitsubishi Heavy Industries, Ltd. (MHI). In the GALAXY-Z code, DSTG can directly treat heterogeneity of dispersed fuel particles in fuel compacts. The ultra-fine-group fixed-source DSTG calculation is applied to high-energy range beyond 30eV, in which up-scattering can be ignored. The fixed-source DSTG calculation is applied to Dancoff factor calculation for the equivalence theory in low-energy range less than 30eV. In the low-energy range, the SHEM 361-group structure is applied, and the energy grids are detailed enough to allow a continuous treatment of the neutron spectrum. In other word, the up-scattering effects and the resonance self-shielding effects are automatically considered in multi-group transport calculation. The homogenized and collapsed multi-group cross-sections of fuel particles are provided to lattice calculations of fuel elements. The comparison of reaction rates (effective cross-sections) between GALAXY-Z and the continuous-energy Monte Carlo code MVP is carried out for fuel particles within a fuel element of the High Temperature Engineering Test Reactor (HTTR). The numerical results indicated that the present method well reproduces fuel particle's reaction rates (effective cross-sections) calculated by MVP.

DOI： 10.13182/PHYSOR24-43462

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 Open Access

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

Open Access

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 Open Access

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

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Papers - YAMAMOTO, Akio