Faculty Profiles - YAMAMOTO, Akio

Papers - YAMAMOTO, Akio

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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 Vol. 214 2025.5

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

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

Oshima, Y; Endo, T; Yamamoto, A

NUCLEAR SCIENCE AND ENGINEERING Vol. 199 ( 4 ) page： 586 - 598 2025.4

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

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 Vol. 62 ( 2 ) page： 179 - 196 2025.2

Language：English Publishing type：Research paper (scientific journal) Publisher：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 UO2 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

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 2024.12

Language：English Publishing type：Research paper (scientific journal) Publisher：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 P3 (SP3) calculation with the local/global 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 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 UO2-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 SP3 calculation. The calculation results show that the proposed method accurately reconstructs the fine mesh distribution of neutron flux.

DOI： 10.1080/00223131.2024.2436958

Flux distribution tallies using proper orthogonal decomposition in Monte Carlo calculations

Kondo, R; Endo, T; Yamamoto, A

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY Vol. 61 ( 12 ) page： 1536 - 1545 2024.12

Language：English Publishing type：Research paper (scientific journal) Publisher：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 Vol. 61 ( 11 ) page： 1405 - 1414 2024.11

Language：English Publishing type：Research paper (scientific journal) Publisher：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 Vol. 302 2024.10

Language：English Publishing type：Research paper (scientific journal) Publisher：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 2024.10

Language：English Publishing type：Research paper (scientific journal) Publisher：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 Vol. 205 2024.9

Language：English Publishing type：Research paper (scientific journal) Publisher：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 2024.8

Language：English Publishing type：Research paper (scientific journal) Publisher：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 2024.7

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 Vol. 61 ( 6 ) page： 830 - 839 2024.6

Language：English Publishing type：Research paper (scientific journal) Publisher：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 Vol. 198 ( 6 ) page： 1215 - 1234 2024.6

Language：English Publishing type：Research paper (scientific journal) Publisher：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 Vol. 198 ( 5 ) page： 981 - 992 2024.5

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

Application of neutron current method for Dancoff factor estimation of fuel particles in double-heterogeneous fuel

Yamamoto, A; Endo, T

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY Vol. 61 ( 3 ) page： 354 - 362 2024.3

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

An evaluation method of Dancoff factors of fuel particles in a typical fuel element of a high-temperature gas-cooled reactor (HTGR) is proposed based on the neutron current method that is widely used for lattice physics calculations of light water reactors. The first level of heterogeneity, i.e. dispersed fuel particles in graphite matrix, is treated by the deterministic statistical geometry (DSTG) method. The homogenized cross-sections of dispersed fuel particles are provided to the second level of heterogeneity, i.e. fuel compacts in a fuel element or a fuel pebble. The validity of the present method is confirmed through the comparison with the reference Dancoff factor obtained by the Monte Carlo and the neutron current methods, which explicitly treats the double heterogeneity. The comparison is carried out for a fuel element simulating the High Temperature Engineering Test Reactor (HTTR), which adopts a typical prismatic fuel element. The numerical results indicate that the present method well reproduces the reference Dancoff factor under various packing fractions. Since the present method can handle flexible geometry and its computation time is short, the present method will be a candidate for the Dancoff factor evaluation method in design calculations of HTGR.

DOI： 10.1080/00223131.2023.2231462

Deterministic sampling method using simplex ensemble and scaling method for efficient and robust uncertainty quantification

Endo, T; Maruyama, S; Yamamoto, A

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY Vol. 61 ( 3 ) page： 363 - 374 2024.3

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

Uncertainty quantification (UQ) of the neutron multiplication factor is important to investigate the appropriate safety margin for a target system. Although the random sampling method is a practical and useful UQ method, a large computational cost is required to reduce the statistical error of the estimated uncertainty. Furthermore, if an input variable follows a normal distribution with a large standard deviation, the perturbed input variable by the random sampling method may become a physically inappropriate or negative value. To address these issues for the efficient and robust UQ, a modified deterministic sampling method using the simplex ensemble and the scaling method is proposed. The features of the proposed method are summarized as follows: The sample size is (Formula presented.), where (Formula presented.) corresponds to the effective rank of the covariance matrix between the input variables; depending on a situation of target UQ, the amounts of perturbations for the input parameters can be arbitrarily given by the scaling factor method; the scaling factor can be updated to avoid physically inappropriate in the perturbed input variables. The effectiveness of the proposed method is demonstrated through the UQ of the neutron multiplication factor due to fuel manufacturing uncertainties for a typical PWR pin-cell burnup calculation.

DOI： 10.1080/00223131.2023.2231931

Uncertainty reduction of sodium void reactivity using data from a sodium shielding experiment Open Access

Maruyama, S; Endo, T; Yamamoto, A

JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY Vol. 61 ( 1 ) page： 31 - 43 2024.1

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

This study investigated the feasibility of reducing the uncertainty associated with fast-reactor-core design by sharing an experimental database between different fields (e.g. reactor physics and radiation shielding) using data assimilation techniques. As the first step in this study, we focused on the ORNL sodium shielding experiment and investigated the possibility of using the experimental data to reduce the uncertainty in sodium void reactivity (SVR), which is the most important safety parameter for sodium-cooled fast reactors. A sensitivity analysis based on the Generalized Perturbation Theory was performed for the sodium shielding experiment. Using the sensitivity coefficients evaluated here and those of the sodium void reactivity previously evaluated by the JAEA, we showed that sodium shielding experimental data can contribute to the uncertainty reduction of SVR by adopting the cross-section adjustment method. Based on this study, the uncertainty reduction effect is expected to be significant, especially for SVR dominated by neutron-leakage phenomena. Although new reactor physics experimental data on SVR may be difficult to obtain, the results of this study suggest that data from sodium shielding experiments can partially substitute for this role. This study demonstrated the value of the mutual use of integral experimental data in fast reactor designs.

DOI： 10.1080/00223131.2023.2244512

Report on the lecture of standard committee in the 2023 fall meeting entitled “Standardization Activities for Safe Long Term Operation”

Murakami Kenta, Onizawa Kunio, Yamamoto Akio

Journal of the Atomic Energy Society of Japan Vol. 66 ( 4 ) page： 199 - 202 2024

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

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

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