研究者詳細 - 山本　章夫

論文 - 山本　章夫

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Hybrid data assimilation methods for nuclear-data-induced uncertainties

Maruyama, S; Yamamoto, A; Endo, T

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 2025年5月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Journal of Nuclear Science and Technology

This study proposes new hybrid data assimilation (DA) methods to effectively use experimental data represented in two different ways in the DA process for reducing nuclear-data-induced uncertainties. Conventional DA methods often assume a linear model, where data in a DA database are represented by their differential coefficients, i.e. nuclear data sensitivity coefficients. In contrast, more rigorous and versatile DA methods based on sampling techniques have been proposed recently. These sampling-based DA methods describe data for DA using results derived from direct transport calculations performed with perturbed nuclear data samples. The proposed hybrid DA methods simultaneously use both types of data described by sensitivity coefficients and sampling results and thus are called ‘hybrid.’ Two hybrid DA methods are proposed herein: the simple hybrid DA and efficient hybrid DA methods. Among these, the simple method is based on a straightforward concept that uses a random sampling technique. In contrast, the efficient method employs a more technical method that effectively combines a new deterministic sampling technique with sensitivity coefficients to reduce statistical uncertainties associated with the simple method. The efficient method is expected to be a candidate to achieve high precision and rigorous DA with a realistic sample size.

DOI： 10.1080/00223131.2025.2500579

Collapsed energy group structure for analysis of prismatic high-temperature gas-cooled reactor with uranium fuel

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

ANNALS OF NUCLEAR ENERGY 214 巻 2025年5月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Annals of Nuclear Energy

Considering computational costs, it is practical to use collapsed energy group structures with less than several tens of groups of the core analysis of High-Temperature Gas-cooled Reactors (HTGRs). This study evaluates the performance of several existing energy group structures and newly optimized 15-group and 25-group structures for prismatic HTGRs with uranium fuel. Both multi-element and two-dimensional core geometries were used to account for neutron spectrum interference effects caused by the presence of control rods, differences in graphite temperature, and variations in the amount of surrounding graphite. The optimization calculations were performed using Genetic algorithm and Particle swarm optimization. Among the group structures with around 15 groups, the newly optimized 15-group structure showed relatively good agreement with the reference solution. For the group structures with around 25 groups, while the newly optimized 25-group structure showed slightly better performance under certain conditions, the existing 26-group structure consistently demonstrated stable and good performance.

DOI： 10.1016/j.anucene.2025.111238

Treatment of corner crossing ray-traces in CMFD acceleration of MOC

Yamamoto, A; Endo, T

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 2025年4月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Journal of Nuclear Science and Technology

Treatment of corner crossing ray traces in the coarse-mesh finite-difference (CMFD) acceleration of a method of characteristics (MOC) transport calculation is discussed. The contribution of the corner crossing ray trace at a CMFD mesh is independently tallied as partial current at the corner of the CMFD mesh to reduce computational burden during transport sweeps in MOC. The tallied partial current at the corner is distributed to the partial currents at the sides of CMFD meshes. Some details of implementations in rectangular and triangular CMFD meshes are explained. The present method is implemented to the GENESIS neutron transport code, and its validity is confirmed through the 3 (Formula presented.) 3 square cells and the hexagonal cell geometries. The calculation results indicate that the present method resolves the issue of corner crossing ray trace in the CMFD acceleration of MOC.

DOI： 10.1080/00223131.2025.2490832

Impact of nuclear data updates from JENDL-4.0 to JENDL-5 on burnup calculations of light-water reactor fuels

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

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 2025年4月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Journal of Nuclear Science and Technology

This study investigated the impact of nuclear data updates from JENDL-4.0 (J4) to JENDL-5 (J5) on the light-water reactor fuel burnup calculations. Burnup calculations were conducted with J4 and J5 for pressurized water reactor pin-cell and boiling water reactor (BWR) fuel assembly geometries. The calculation results revealed significant burnup-dependent differences in the neutron multiplication factor (k<inf>inf</inf>). Across the burnup range of 0–50 GWd/t, k<inf>inf</inf> values of J5 were consistently smaller than those of J4 and the difference gradually increased as burnup progressed. Direct sensitivity calculations, in which each nuclide data was replaced from J4 to J5, indicated that updates to the cross-sections of U-235, U-238, and Pu-239 and the thermal scattering law data of H in H<inf>2</inf>O notably impacted the k<inf>inf</inf> differences. For the BWR assembly geometry containing Gd fuels, large k<inf>inf</inf> differences were observed in the burnup range of 10–15 GWd/t. This difference was primarily attributed to updates in the cross-sections of U-235, Gd-155, Gd-157, and thermal scattering law data of H in H<inf>2</inf>O. Furthermore, we investigated how the nuclear data updates affected the k<inf>inf</inf> differences by examining nuclide number densities, the energy-dependent sensitivities, and the neutron spectra.

DOI： 10.1080/00223131.2025.2486798

Development of Multigroup Monte Carlo Neutron Transport Method with Regionwise Even-Parity Discontinuity Factor

Oshima, Y; Endo, T; Yamamoto, A

NUCLEAR SCIENCE AND ENGINEERING 199 巻 ( 4 ) 頁： 586 - 598 2025年4月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Nuclear Science and Engineering

The multigroup Monte Carlo (MC) neutron transport method with a regionwise even-parity discontinuity factor (REPDF), i.e. the discontinuity factor (DF)–MC method, is developed with the aim to provide a reference solution for deterministic transport calculations with DF. Applying the analogy with optics, neutrons are transmitted or reflected at a region surface during random walks. The probability of transmission or reflection is determined by REPDFs in adjacent regions. The DF is traditionally used in deterministic neutron transport methods to reduce the discretization error due to spatial homogenization and energy condensation. The DF-MC method can treat DF in the framework of the multigroup MC method. In this paper, the weight cancellation technique based on the closest pair of points using the divide-and-conquer algorithm is used because negative weights appear due to the neutron reflection. The REPDF is calculated by the method of characteristics (MOC). The verification calculations are carried out in the pin-by-pin homogenized and assembly homogenized KAIST-2A core geometry. The DF-MC calculation can reproduce the results of the MOC with the REPDF. These results demonstrate the principle of the DF-MC method and extend the application of the DF to the probabilistic neutron transport method.

DOI： 10.1080/00295639.2024.2383102

Reduced order SP<sub>3</sub> core calculation based on local/global iterations using proper orthogonal decomposition

Ito, M; Endo, T; Yamamoto, A; Takeishi, T; Kodama, Y; Hyoudou, H

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 62 巻 ( 4 ) 頁： 366 - 378 2025年4月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Journal of Nuclear Science and Technology

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 simplified P<inf>3</inf> (SP<inf>3</inf>) calculation with the local/global iterations. The fine mesh SP<inf>3</inf> calculation (local calculation) for each single assembly can be efficiently solved using POD, thanks to the matrix form-based SP<inf>3</inf> equation and the relatively small system of linear 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 partial current-based coarse mesh finite difference (p-CMFD) calculation is applied to the coarse mesh calculation (global calculation). Using the global calculation results, each local boundary condition can be updated. These local and global calculations are iterated until convergence. Through a two-dimensional UO<inf>2</inf>-MOX whole core calculation with different fuel loading patterns, the proposed method is verified in comparison with the reference result by the conventional fine mesh SP<inf>3</inf> calculation. The calculation results show that the proposed method accurately reconstructs the fine mesh distribution of neutron flux.

DOI： 10.1080/00223131.2024.2436958

Verification of direct coupling code system using FRENDY version 2 and GENESIS for light water reactor lattices

Fujita, T; Yamamoto, A

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 62 巻 ( 2 ) 頁： 179 - 196 2025年2月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Journal of Nuclear Science and Technology

This study newly established a direct coupling code system consisting of the nuclear data processing code FRENDY version 2, and the three-dimensional heterogeneous transport code GENESIS (FRENDY-V2/GENESIS) for easy implementation of the random-sampling-based uncertainty quantification considering the implicit effect due to nuclear cross-section (XS) perturbations. The multi-group macroscopic XSs prepared for GENESIS were generated by FRENDY version 2, where the Dancoff factor was calculated by the neutron current method. Then the background XSs were evaluated based on the Carlvik two-term rational approximation. The infinite multiplication factor (k-infinity) and the fission reaction rate distribution in UO<inf>2</inf> and MOX lattice geometries were compared with MVP3 to verify the calculation accuracy of FRENDY-V2/GENESIS. The sensitivity analyses on the discretization conditions such as the ray tracing of the method of characteristics were also carried out. Through several comparisons between FRENDY-V2/GENESIS and MVP3, FRENDY-V2/GENESIS with the SHEM 361-group structure calculates the k-infinity within approximately 50 pcm and the fission reaction rate distribution within approximately 0.1% by the root mean square, respectively. Consequently, the applicability of FRENDY-V2/GENESIS was verified, and FRENDY-V2/GENESIS can be used to discuss the implicit effect due to multi-group XS perturbations.

DOI： 10.1080/00223131.2024.2410918

将来世代のための原子力の平和利用の3Sに関する討議 1

山本章夫

日本原子力学会誌ATOMOΣ 67 巻 ( 4 ) 頁： 218 - 221 2025年

記述言語：日本語掲載種別：研究論文（学術雑誌）出版者・発行元：一般社団法人日本原子力学会

<p>　Safety, Security, Safeguards(3S)の一つである原子力安全(Safety)の概念について，(1)原子力安全の目的，(2)安全目標，(3)深層防護，(4)安全設計・安全評価，(5)「欠け」と想定外事象対応を解説する。また，Safety, Security, Safeguardsはお互いに関連するため，Safetyの観点から相補的あるいは相反的な点について留意事項を説明する。本稿は2024年日本原子力学会秋の大会における核不拡散・保障措置・核セキュリティ連絡会と学生連絡会との合同企画セッションの内容がベースになっており，この分野にあまりなじみのない「将来世代」の読者を想定している。</p>

DOI： 10.3327/jaesjb.67.4_218

Fundamental properties and characteristics of flux distribution tallies using proper orthogonal decomposition

Kondo R., Yamamoto A., Endo T.

EPJ Nuclear Sciences and Technologies 11 巻 2025年

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：EPJ Nuclear Sciences and Technologies

The flux distribution tallies using the proper orthogonal decomposition (POD) called the POD tallies have been developed in our previous study. The POD tallies can achieve dimensionality and statistical uncertainty reduction for a finely discretized flux distribution. Some characteristics of the POD tallies, which are left by our previous work, are revealed in the present study. Firstly, the POD tallies with the track length estimator are newly implemented. Since the statistical uncertainty of the POD tallies is reduced compared with the cell tallies, the POD tallies with the track length estimator can obtain the most precise result among the present implantations. Secondly, the basis vectors obtained by the deterministic and the stochastic methods are compared. The statistical uncertainty of the snapshot data invokes the degradation of the extracted basis vectors. This result indicates that the deterministic method might be more efficient for the snapshot calculation. Finally, the impact of the covariances of expansion coefficients on the statistical uncertainty of expanded flux distribution is investigated. The reconstructed statistical uncertainty considering only the variances of the expansion coefficients differs from the reference. This result reveals that the covariances of the expansion coefficients are important to estimate the statistical uncertainty of the local flux in the flux distribution.

DOI： 10.1051/epjn/2025019

A deterministic transport calculation method of α-eigenvalue for non-multiplying system using framework of k-eigenvalue calculation

Yamamoto A., Tsujita K., Endo T.

Journal of Nuclear Science and Technology 2025年

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Journal of Nuclear Science and Technology

A new α-eigenvalue calculation method for a non-multiplying system using the framework of conventional k-eigenvalue calculation of a deterministic transport code is proposed. The time-derivative or the time-absorption term in α-eigenvalue calculation is moved to the right-hand side of the transport equation and is considered as a virtual fission source. The present method is applied to a simple three-dimensional problem with the one-group cross-section and a small water tank system with the 252-group cross-section. The calculation results indicate that the present method provides accurate α-eigenvalue with rapid convergence and no numerical instability.

DOI： 10.1080/00223131.2025.2507866

Flux distribution tallies using proper orthogonal decomposition in Monte Carlo calculations

Kondo, R; Endo, T; Yamamoto, A

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 61 巻 ( 12 ) 頁： 1536 - 1545 2024年12月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Journal of Nuclear Science and Technology

In this paper, a new tallying method for a neutron flux distribution using the proper orthogonal decomposition is proposed for dimensionality reduction. The target spatial flux distribution is expanded by orthogonal basis vectors. Expansion coefficients are tallied during the random walk of the Monte Carlo calculation. The orthogonal basis vectors are extracted from the pre-calculated snapshots by the singular value decomposition. The proposed method is verified in the multi-group Monte Carlo calculation with the one-dimensional heterogeneous whole core geometry as a feasibility study. The flux distribution for each of the assemblies and energy groups is expanded by the basis vectors. The fewer basis vectors obtained from snapshots can reconstruct the target distribution well compared with the conventional Legendre polynomials used in the functional expansion tallies. The dimension of the solution in the proposed method is reduced by a factor of twenty compared with the conventional cell tally. In addition, the statistical error is reduced through dimensionality reduction thanks to the methodological feature of the proposed method. The results indicate that the proposed method has the capability of dimensionality reduction to tally the finely discretized flux distribution.

DOI： 10.1080/00223131.2024.2365445

Limited linear source approximation with edge detection for convergence stability of method of characteristics

Yamamoto, A; Endo, T; Chiba, G

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 61 巻 ( 11 ) 頁： 1405 - 1414 2024年11月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Journal of Nuclear Science and Technology

A new implementation of the limited linear source approximation (LLSA) is proposed. The LLSA was previously proposed to eliminate local negative source in flux regions to mitigate numerical instability of the method of characteristics (MOC) with linear source approximation (LSA). In the present LLSA implementation, the convex edges of flux regions are used to check the local negative source to decrease the computational load. The present method is implemented in the transport code GENESIS, and its effectiveness is verified through the two dimensional C5G7 benchmark problem and the simplified two-dimensional high-temperature engineering test reactor core. The calculation results indicate that the present LLSA implementation efficiently mitigates the numerical instability of MOC with LSA. Additional computational time is less than 1% of total computation time.

DOI： 10.1080/00223131.2024.2341776

Implementation of track length estimator for flux distribution tallies using proper orthogonal decomposition in one-dimensional geometry Open Access

Kondo R., Endo T., Yamamoto A.

EPJ Web of Conferences 302 巻 2024年10月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：EPJ Web of Conferences

In a recent study, we have developed an efficient flux distribution tallying method in the Monte Carlo calculation toward the high-fidelity, large scale multi-physics simulation. In this method, the proper orthogonal decomposition is applied to the flux distribution tallies. While the tallying method was implemented with the collision estimator in the previous study, the track length estimator is implemented in the present study to obtain the tally with lower statistical error. The implementation of the flux distribution tally with the track length estimator is compared with that of the collision estimator and the normal track length estimator in a one-dimensional problem. The numerical results reveal that the distribution tally using the proper orthogonal decomposition with the track length estimator can obtain a more precise solution compared with that with the collision estimator. Therefore, in terms of the statistical error, the relationship between the distribution tally with track length and collision estimators is similar to that between the conventional track length and collision estimators.

DOI： 10.1051/epjconf/202430204002

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

NUCLEAR SCIENCE AND ENGINEERING 2024年10月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Nuclear Science and Engineering

A resonance calculation using a deterministic statistical geometry (DSTG) method has been developed to efficiently treat the double heterogeneity of coated fuel particles loaded in a high-temperature gas-cooled reactor. Using the DSTG method, neutron flux spatial distribution can be calculated in a heterogeneous geometry containing the randomly dispersed fuel particles and surrounding graphite matrix, which is consistent with statistical geometry 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. In the GALAXY-Z code, DSTG can directly treat the heterogeneity of dispersed fuel particles in fuel compacts. The ultra-fine-group fixed-source DSTG calculation is applied to a high-energy range beyond 30 eV in which upscattering can be ignored. The fixed-source DSTG calculation is applied to a Dancoff-factor calculation for the equivalence theory in a low-energy range of less than 30 eV. 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 words, the upscattering effects and the resonance self-shielding effects are automatically considered in the multigroup transport calculation. The homogenized and collapsed multigroup 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. The numerical results indicate that the present method well reproduces the fuel particle’s effective cross sections, k-infinity, and fission rate distribution calculated by MVP.

DOI： 10.1080/00295639.2024.2397259

Quantifying uncertainty induced by scattering angle distribution using maximum entropy method

Maruyama, S; Yamamoto, A; Endo, T

ANNALS OF NUCLEAR ENERGY 205 巻 2024年9月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Annals of Nuclear Energy

This study developed a new method for evaluating the uncertainty in reactor core/shielding characteristics attributable to the scattering angle distribution in the combined method of the continuous energy Monte Carlo (CEMC) transport calculations and the random sampling (RS) technique based on nuclear data covariances. Recent advances in computer performance have made it possible to evaluate the nuclear data-induced uncertainty with high accuracy using the Monte Carlo method. The total Monte Carlo (TMC) method is a typical uncertainty quantification method that relies entirely on the Monte Carlo method. Another representative Monte Carlo-based uncertainty quantification method is the combination method of the RS technique based on nuclear data covariances and CEMC transport calculations (CEMC-RS). CEMC-RS has the advantage that the uncertainty can be quantified even if one is not familiar with nuclear data measurement and evaluation, as long as the nuclear data covariances are available. However, no method has been established to quantify the uncertainty in CEMC-RS, despite the fact that the uncertainty due to the scattering angle distribution is non-negligible in fast reactor core analysis and shielding analysis. This study introduces a new approach for uncertainty quantification related to scattering angle distribution in CEMC-RS, utilizing the maximum entropy method. The effectiveness of this method was verified through comparison with results from the classical deterministic uncertainty quantification approach based on generalized perturbation theory. Overall, this method offers a more accurate tool for nuclear engineers and researchers in evaluating and managing uncertainties in reactor design and safety analysis.

DOI： 10.1016/j.anucene.2024.110591

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

NUCLEAR SCIENCE AND ENGINEERING 2024年8月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Nuclear Science and Engineering

An improved sampling method for flight distance is proposed for Monte Carlo analysis of TRISO fuel particles using the statistical geometry (STG) method. The statistically uniform distribution of fuel particles, which is usually assumed as a default sampling method of flight distance of a neutron between fuel particles, shows considerable bias on k-infinity when coating layers of a TRISO fuel particle are homogenized with a graphite matrix. The proposed new sampling method almost resolves the difference between the no-coating-layer model (coating layers are homogenized with a graphite matrix) and the explicit-coating-layer model (explicitly considers coating layers, reference). By adopting the present method, the no-coating-layer model can be used without significant loss of accuracy in the STG method of a Monte Carlo analysis. The computation time with a continuous energy Monte Carlo code for a typical fuel compact cell of a high-temperature gas-cooled reactor is reduced to one-seventh of the explicit-coating-layer model when the no-coating-layer model is used.

DOI： 10.1080/00295639.2024.2384236

Uncertainty reduction of sodium void reactivity using data from a sodium shielding experiment (vol 61, pg 31, 2024)

Maruyama, S; Endo, T; Yamamoto, A

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 2024年7月

Development of nuclear data processing code FRENDY version 2

Tada, K; Yamamoto, A; Kunieda, S; Konno, C; Kondo, R; Endo, T; Chiba, G; Ono, M; Tojo, M

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY 61 巻 ( 6 ) 頁： 830 - 839 2024年6月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Journal of Nuclear Science and Technology

Nuclear data processing is important to connect evaluated nuclear data libraries and radiation transport codes. The nuclear data processing code FRENDY version 1 was released in 2019 to generate ACE-formatted cross section files with simple input data. After we released FRENDY version 1, many functions were developed, e.g. neutron multi-group cross section generation, explicit consideration of the resonance interference effect among different nuclides in a material, adaptive setting of the background cross sections, consideration of the resonance upscattering, ACE file perturbation, statistical uncertainty quantification of probability table, and modification of ENDF-6-formatted file. FRENDY version 2 was released including these new functions. It generates GENDF- and MATXS-formatted neutron multi-group cross section files from an ACE-formatted cross section file or an evaluated nuclear data file. This paper explains the features of the new functions implemented in FRENDY version 2 and the verification of the neutron multigroup cross section generation function of this code.

DOI： 10.1080/00223131.2023.2278600

Impact of Uncertainty Reduction on Lead-Bismuth Coolant in Accelerator-Driven System Using Sample Reactivity Experiments

Katano, R; Oizumi, A; Fukushima, M; Pyeon, CH; Yamamoto, A; Endo, T

NUCLEAR SCIENCE AND ENGINEERING 198 巻 ( 6 ) 頁： 1215 - 1234 2024年6月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Nuclear Science and Engineering

In this study, we have demonstrated that data assimilation (DA) using lead and bismuth sample reactivities measured in the Kyoto University Critical Assembly A-core can successfully reduce the uncertainty of the coolant void reactivity in accelerator-driven systems (ADSs) derived from inelastic scattering cross sections of lead and bismuth. We reevaluated and highlighted the experimental uncertainties and correlations of the sample reactivities for the DA formula. We used the MCNP6.2 code to evaluate the sample reactivities and their uncertainties and performed DA using the reactor analysis code system MARBLE. The high-sensitivity coefficients of the sample reactivities to lead and bismuth allowed us to reduce the cross-section–induced uncertainty of the void reactivity of the ADS from 6.3% to 4.8%, achieving a provisional target accuracy of 5% in this study. Furthermore, we demonstrated that the uncertainties arising from other dominant factors, such as minor actinides and steel, can be effectively reduced by using integral experimental data sets for the unified cross-section dataset ADJ2017.

DOI： 10.1080/00295639.2023.2246779

Deterministic Transport Calculation Method for Statistical Geometry with Small Fuel Particles

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

NUCLEAR SCIENCE AND ENGINEERING 198 巻 ( 5 ) 頁： 981 - 992 2024年5月

記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Nuclear Science and Engineering

A deterministic transport calculation method is proposed for the treatment of dispersed fuel particles in a fuel compact/fuel pebble of a typical high-temperature gas-cooled reactor fuel. The random distribution of fuel particles was considered using the statistical geometry (STG) method, which is widely used in the Monte Carlo method. A long-ray trace, which represents a neutron flight path, was considered, and the segment lengths and material distributions on the ray trace were randomly sampled using STG. Then a conventional transport sweep, as used in the method of characteristics, was performed along the ray trace. The proposed deterministic statistical geometry (DSTG) method can calculate the flux spatial distribution in a heterogeneous geometry containing randomly dispersed fuel particles and the surrounding graphite matrix, which is consistent with the STG in a Monte Carlo method. The validity of the DSTG method was confirmed through sensitivity calculations and comparisons with a multigroup Monte Carlo method that utilizes STG. The proposed method can be used for the homogenization of heterogeneous structures inside a fuel compact or fuel pebble as an alternative to conventional deterministic unit cell calculations that consider fuel particles and the surrounding matrix in high-temperature gas-cooled reactor fuels.

DOI： 10.1080/00295639.2023.2230414

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