Faculty Profiles - YAMAMOTO, Akio

Papers - YAMAMOTO, Akio

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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

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

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

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

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

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

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

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

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

ACE-FRENDY-CBZ: a new neutronics analysis sequence using multi-group neutron transport calculations

Chiba, G; Yamamoto, A; Tada, K

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY Vol. 60 ( 2 ) page： 132 - 139 2023.2

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

We propose a new neutronics analysis sequence using multi-group neutron transport calculations named ACE-FRENDY-CBZ. This sequence is free from uses of any application libraries; with the ACE files as the starting point, multi-group cross sections of media comprising a target system are calculated with the FRENDY code, and multi-group neutron transport calculations are performed with modules of the CBZ code system. The ACE-FRENDY-CBZ sequence was tested against the eight fast neutron systems, and good agreement in the neutron multiplication factors with the reference Monte Carlo results was obtained within 20 pcm differences in the bare systems and within 60 pcm differences in the reflected systems. It was also found that the adoption of the consistent P approximation increases the errors. In order to investigate this issue, we adopted the sub-group method to calculate spatially-dependent current-weighted total cross sections in the reflector regions, and it was suggested that the uses of the spatially-dependent cross sections with the consistent P approximation has a possibility to further improve the numerical accuracy.

DOI： 10.1080/00223131.2022.2087783

Application of Equivalent Dancoff Factor Method for Resonance Calculation of Double Heterogeneous Fuel

Yamamoto A., Endo T.

Transactions of the American Nuclear Society Vol. 128 page： 690 - 693 2023

Roadmap towards the introduction of new design fuels as well as the continuous improvement of safety evaluation (5)

Murakami Kenta, Yamamoto Akio

Journal of the Atomic Energy Society of Japan Vol. 65 ( 10 ) page： 606 - 607 2023

Implementation of Resonance Up-scattering Treatment in FRENDY Nuclear Data Processing System

Yamamoto, A; Endo, T; Chiba, G; Tada, K

NUCLEAR SCIENCE AND ENGINEERING Vol. 196 ( 11 ) page： 1267 - 1279 2022.11

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

The resonance upscattering effect (the thermal agitation effect) is implemented in the generation capability of multigroup neutron cross sections of the FRENDY nuclear data processing system. The resonance upscattering effect is considered by (1) the variation of self-shielding factors (effective cross sections) due to the change in the ultra-fine group spectrum and (2) the variation of group-to-group elastic scattering cross sections. Since the upscattering effect is considered in the ultra-fine group spectrum calculation, an iteration calculation is necessary to consider the upscattering. The impacts of the iteration strategy (Jacobi or Gauss-Seidel), as well as the number of iterations, are discussed. In the verification calculations, impacts on the ultra-fine group spectrum, effective cross sections, and neutronics characteristics (the Doppler effect) are confirmed. The effect of energy group structure and the impact of resonance upscattering treatments on the Doppler effect through the variation of effective cross sections and the elastic scattering matrix are investigated. The results indicate that FRENDY can provide appropriate multigroup cross sections considering the resonance upscattering effect.

DOI： 10.1080/00295639.2022.2087833

Sensitivity Coefficient Evaluation of an Accelerator-Driven System Using ROM-Lasso Method

Katano, R; Yamamoto, A; Endo, T

NUCLEAR SCIENCE AND ENGINEERING Vol. 196 ( 10 ) page： 1194 - 1208 2022.10

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

We propose the use of reduced-order modeling to improve the sensitivity coefficient evaluation method based on Lasso-type penalized linear regression. In this method, cross sections of interest are uniformly randomly sampled, and corresponding perturbed core analyses are performed. The sensitivity coefficients of the higher-dimensional model are expanded by the active subspace (AS) attained by the lower-dimensional model, and the expansion coefficients are estimated by the Lasso regression. In addition, AS bases can be flexibly chosen according to neutronics parameters of interest. We conducted a verification calculation for an accelerator-driven system and clarified that the proposed method successfully reduces the calculation cost by a couple of orders of magnitude compared with the direct method. The proposed method can be used to practically evaluate the sensitivity coefficients of various parameters.

DOI： 10.1080/00295639.2022.2067447

Application of dynamic mode decomposition to Rossi-α method in a critical state using file-by-file moving block bootstrap method

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

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY Vol. 59 ( 9 ) page： 1117 - 1126 2022.9

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

Prompt neutron decay constant (Formula presented.) in a critical state is useful information to validate the numerically predicted ratio of the point kinetics parameters (Formula presented.), where (Formula presented.) and (Formula presented.) are the effective delayed neutron fraction and prompt neutron lifetime, respectively. To directly measure (Formula presented.) in a target critical system, this study proposes the application of the dynamic mode decomposition (DMD) to the reactor noise analysis based on the Rossi- (Formula presented.) method. The DMD-based Rossi- (Formula presented.) method enables us to robustly estimate the fundamental mode component of (Formula presented.) from the Rossi- (Formula presented.) histograms measured using multiple neutron detectors. Furthermore, the file-by-file moving block bootstrap method is newly proposed for the statistical uncertainty quantification of (Formula presented.) to prevent huge memory usage when the neutron count rate is high and/or the total measurement time is long. A critical experiment has been conducted at Kyoto University Critical Assembly to demonstrate the proposed method. As a result, the proposed method can uniquely determine the (Formula presented.) value of which the statistical uncertainty is smallest. By utilizing this experimental result of (Formula presented.), numerical results of (Formula presented.) ratio using the continuous energy Monte Carlo code MCNP6.2 with recent nuclear data libraries, which are processed by the nuclear data processing code FRENDY, are validated.

DOI： 10.1080/00223131.2022.2030260

Improvements in Computational Efficiency for Resonance Calculation Using Energy Spectrum Expansion Method

Kondo, R; Endo, T; Yamamoto, A; Takeda, S; Koike, H; Yamaji, K; Asano, K

NUCLEAR SCIENCE AND ENGINEERING Vol. 196 ( 7 ) page： 769 - 791 2022.7

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

Improvements in computational efficiency for the Resonance calculation using energy Spectrum Expansion (RSE) method are proposed in order to increase the applicability of the method for core nuclear analyses. First, efficient treatment of the neutron source for the RSE method has been newly developed. This is a balanced approach from the viewpoints of computation time and memory size, in comparison with the other approaches mentioned in a previous study [R. KONDO et al., “A New Resonance Calculation Method Using Energy Expansion Based on a Reduced Order Model,” Nucl. Sci. Eng., 195, 694 (2021)]. Second, low-rank approximation has been applied to the RSE method considering the deficit ratio of the singular value for the orthogonal basis. Computation time was reduced by ~68% while maintaining sufficient accuracy of effective cross sections. Third, the impacts of the discretization parameters in the method of characteristics on the RSE method have been investigated, and coarser conditions of the parameters were found to be appropriate from the viewpoints of computation time and accuracy of effective cross sections. Finally, RSE calculations with these improvements have been performed for the fuel assembly geometry of a light water reactor. The computation time was reduced by ~70%, and the data size of the scattering cross-section moments was approximately 3900 times smaller in comparison with the RSE calculation without the improvements.

DOI： 10.1080/00295639.2021.2025297

Impact of Angular and Spatial Source Distribution Approximations on Convergence Performance of Nonlinear Acceleration Methods for MOC in Slab Geometry

Oshima, Y; Endo, T; Yamamoto, A

NUCLEAR SCIENCE AND ENGINEERING Vol. 196 ( 4 ) page： 379 - 394 2022.4

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

The convergence performance of nonlinear acceleration methods for the method of characteristics (MOC) with flat source (FS) approximation (FS MOC) or linear source (LS) approximation (LS MOC) is numerically investigated by focusing on the spatial and angular approximations in the acceleration calculations. The convergence of nonlinear acceleration depends on the consistency of the calculation models between the higher-order and lower-order (acceleration) methods. The convergence of four acceleration methods is evaluated to clarify the relationship between model consistency and convergence performance. These methods consist of FS or LS for the spatial source distribution and P1 or discrete angle for the angular distribution, i.e., (1) FS analytic coarse mesh finite difference (ACMFD) acceleration (FS ACMFD), (2) LS ACMFD, (3) FS angular-dependent discontinuity factor MOC (ADMOC) acceleration (FS ADMOC), and (4) LS ADMOC. The ACMFD and ADMOC accelerations are based on P1 and discrete angle approximations, respectively. The FS MOC and LS MOC are considered higher-order methods. The FS MOC and LS MOC with five acceleration methods, i.e., the aforementioned four acceleration methods and the conventional coarse mesh finite difference acceleration method, are used to perform fixed-source calculations in one-group one-dimensional homogeneous slab geometry, and the spectral radii are numerically evaluated. The numerical results indicate that (1) the nonlinear acceleration methods that are unconditionally stable for FS MOC also show similar convergence properties for LS MOC in one-dimensional slab geometry; (2) better convergence is observed when the consistency of higher- and lower-order models is high; and (3) when a coarse mesh is optically thick, the spatial homogenization degrades the convergence performance, even if spatial and angular approximations are consistent between the higher- and lower-order models.

DOI： 10.1080/00295639.2021.1982549

Transport consistent diffusion coefficient for CMFD acceleration and comparison of convergence properties (vol 56, pg 716, 2019)

Yamamoto, A; Endo, T; Giho, A

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY Vol. 59 ( 3 ) page： 407 - 407 2022.3

Applicability of Dynamic Mode Decomposition to Estimate Fundamental Mode Component of Prompt Neutron Decay Constant from Experimental Data

Nishioka, F; Endo, T; Yamamoto, A; Yamanaka, M; Pyeon, CH

NUCLEAR SCIENCE AND ENGINEERING Vol. 196 ( 2 ) page： 133 - 143 2022.2

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

To robustly estimate the fundamental mode component of prompt neutron decay constant α in a subcritical system, dynamic mode decomposition (DMD) is applied to time-series data obtained by the pulsed-neutron source (PNS) and Rossi-α methods. For the statistical uncertainty quantification of α by DMD, randomly sampled virtual data are used for the DMD procedure. The applicability of DMD is demonstrated by analyzing the experimental results by the PNS and Rossi-α methods, which are performed at the Kyoto University Critical Assembly (KUCA). When applying the DMD to the PNS and Rossi-α experimental data, a constant signal was added to the experimental data to remove the background constant component. The application results indicate that DMD enables one to robustly estimate the fundamental mode component of α in the PNS and Rossi-α methods.

DOI： 10.1080/00295639.2021.1968225

Safety measures and risk management of nuclear power plants against external events (4)

Murakami Kenta, Yamamoto Akio

Journal of the Atomic Energy Society of Japan Vol. 64 ( 5 ) page： 272 - 274 2022

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