Faculty Profiles - ENDO Tomohiro

Papers - ENDO Tomohiro

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A New Interpretation of Discontinuity Factor Reviewed International journal

Akio Yamamoto, Tomohiro Endo

Nuclear Science and Engineering Vol. 193 ( 9 ) page： 991 - 997 2019.9

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

A new interpretation of the discontinuity factor (DF) for scalar flux, partial current, and angular flux is discussed. Conventionally, the DF is considered as the discontinuous condition of scalar flux, partial current, or angular flux at an interface. In the new interpretation, the DF is considered as the refractive index of materials for partial current or angular flux that conserves odd-parity or odd-moment angular flux at an interface of different materials. It is related to the transmission and reflection of partial current or angular flux at an interface where different materials are adjacent. Using the present interpretation, a fundamental issue of neutron balance (i.e., artificial loss or production of neutrons at an interface due to discontinuous condition), which would appear in the conventional interpretation of DF, can be resolved.

DOI： 10.1080/00295639.2019.1579514

Web of Science

First nuclear transmutation of Np237 and Am241 by accelerator-driven system at Kyoto University Critical Assembly Reviewed International journal

Cheol Ho Pyeon, Masao Yamanaka, Akito Oizumi, Masahiro Fukushima, Go Chiba, Kenichi Watanabe, Tomohiro Endo, Wilfred G. Van Rooijen, Kengo Hashimoto, Atsushi Sakon, Naoto Aizawa, Yasutoshi Kuriyama, Tomonori Uesugi, Yoshihiro Ishi

Journal of Nuclear Science and Technology Vol. 56 ( 8 ) page： 684 - 689 2019.8

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

This study demonstrates, for the first time, the principle of nuclear transmutation of minor actinide (MA) by the accelerator-driven system (ADS) through the injection of high-energy neutrons into the subcritical core at the Kyoto University Critical Assembly. The main objective of the experiments is to confirm fission reactions of neptunium-237 (Np237) and americium-241 (Am241), and capture reactions of Np237. Subcritical irradiation of Np237 and Am241 foils is conducted in a hard spectrum core with the use of the back-to-back fission chamber that obtains simultaneously two signals from specially installed test (Np237 or Am241) and reference (uranium-235) foils. The first nuclear transmutation of Np237 and Am241 by ADS soundly implemented by combining the subcritical core and the 100 MeV proton accelerator, and the use of a lead-bismuth target, is conclusively demonstrated through the experimental results of fission and capture reaction events.

DOI： 10.1080/00223131.2019.1618406

Web of Science

Transport consistent diffusion coefficient for CMFD acceleration and comparison of convergence properties Reviewed International journal

Akio Yamamoto, Tomohiro Endo, Akinori Giho

Journal of Nuclear Science and Technology Vol. 56 ( 8 ) page： 716 - 723 2019.8

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

A diffusion coefficient for the coarse mesh finite difference (CMFD) acceleration is derived from the semi-analytic solution of one-group, one-dimensional, even-parity transport equation. The derived diffusion coefficient, i.e., the transport consistent diffusion coefficient (TCD), depends on the optical length of a mesh and shows similar behavior with the artificial grid diffusion (AGD) and the effective diffusion (EffD) coefficients for an optically thick mesh. Convergence properties of typical diffusion coefficients are evaluated using the linearized Fourier analysis. Analyses of the C5G7 3D benchmark problems with and without voided region are carried out to compare the convergence properties. The number of transport sweeps to reach convergence using TCD is smaller than that using EffD or AGD.

DOI： 10.1080/00223131.2019.1618405

Web of Science

A simple treatment of increased gap due to fuel assembly bowing through correction of cross sections Reviewed International journal

Akio Yamamoto, Tomohiro Endo, Hiroaki Nagano, Yasunori Ohoka, Kento Yamamoto

Journal of Nuclear Science and Technology Vol. 56 ( 6 ) page： 471 - 478 2019.6

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

Increased fuel assembly gap due to bowing in commercial light-water reactors (LWRs) has an impact on local pin-power distribution due to increased local moderation. In order to consider the effect of increased assembly gap without explicit consideration of increased gap width, a correction method of cross sections in the gap region is proposed. In this cross section correction method, the average chord length of gap region is preserved to capture the effect of increased gap width. The validity of the present method is confirmed by verification calculations in a single assembly and 5 × 5 assembly geometries using the GENESIS code, which is a transport code based on the method of characteristics.

DOI： 10.1080/00223131.2019.1598509

Web of Science

Utilization of Regionwise Even-Parity Discontinuity Factor to Reduce Discretization Error of MOC Reviewed International journal

Yamamoto Akio, Giho Akinori, Endo Tomohiro

Nuclear Science and Engineering Vol. 193 ( 3 ) page： 253 - 268 2019.3

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

To reduce angular and spatial discretization error of the method of characteristics with a coarse calculation condition, the regionwise even-parity discontinuity factor (EPDF) for transport calculations is evaluated through an iterative procedure using only the regionwise scalar flux, i.e., without the odd-parity angular flux, the partial current, or the net current at the region boundary. The regionwise EPDF is evaluated in a single-assembly geometry with the reflective boundary condition. The evaluated EPDF is applied to a 2 × 2 colorset assembly and core configurations, and the performance is compared to that of the conventional superhomogenization (SPH) method. The calculation results indicate that (1) no convergence issue is observed during the iteration process to estimate the EPDF, (2) the performance of the regionwise EPDF is better than that of the conventional SPH method, and (3) the normalization of the EPDF is necessary to incorporate different surface scalar flux levels among different types of fuel assemblies.

DOI： 10.1080/00295639.2018.1516961

Web of Science

Inverse estimation methods of unknown radioactive source for fuel debris search Invited Reviewed International journal

Shinji Sugaya, Tomohiro Endo, Akio Yamamoto

Annals of Nuclear Energy Vol. 124 page： 49 - 57 2019.2

Language：English Publishing type：Research paper (scientific journal) Publisher：Elsevier B.V.

To identify the distribution of fuel debris remaining in the reactor vessel and/or the containment vessel of Fukushima Daiichi NPS, we focused on the inverse estimation of radioactive source distribution using the measured radiation counts. The Maximum Likelihood-Expectation Maximization (ML-EM) and the Moore-Penrose Matrix Inverse (MPMI) methods are examined. The ML-EM method has been used for the image reconstruction of computed tomography, and the MPMI method is one of the solution methods for simultaneous linear equations with the underdetermined condition. A simple calculation model simulating a containment vessel was constructed including detectors and radiation sources. In an actual situation, a sufficient number of radiation measurement positions would not be available owing to the complexity of structures inside the containment vessel. Thus, the number of radiation measurement points (number of constraints) is smaller than that of radiation source positions. It means that an underdetermined inverse problem should be solved. The detection probability of radiation (neutron or photon) is calculated by the adjoint transport calculation since the detection probability is used as the coupling coefficient between radiation counts at a detector and a radioactive source. The result of estimation using the ML-EM or the MPMI method indicates that the accuracy of estimation depends on the distance between a radiation source and a detector, and measurement positions of radiation count. The ML-EM and the MPMI methods show different prediction accuracy depending on the prediction condition. It is found that reasonable prediction accuracy would be obtained when the detectors are placed at the vicinity of radiation sources of interest.

DOI： 10.1016/j.anucene.2018.09.022

Flux Region Assignment Method Using Ray Trace Information for the Method of Characteristics to Improve Cache Efficiency Reviewed International journal

Akio Yamamoto, Akinori Giho, Tomohiro Endo

Nuclear Science and Engineering Vol. 192 ( 3 ) page： 240 - 253 2018.12

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

A flux region assignment algorithm to increase cache efficiency for the method of characteristics (MOC) is proposed. In order to minimize the stride of memory access, flux region identifications are assigned based on the ray trace sequence during the MOC calculation. The present method is implemented in the three-dimensional transport code GENESIS and its performance is confirmed through verification calculations ranging from single pressurized water reactor (PWR) fuel assembly to PWR full-core benchmark problems. Quantitative comparison of cache efficiency is carried out and the present method shows improved cache efficiency, which results in a reduction in computation time. The present method can reduce computational time by improving cache efficiency while suppressing memory requirement.

DOI： 10.1080/00295639.2018.1501978

Web of Science

Underestimation of statistical uncertainty of local tallies in Monte Carlo eigenvalue calculation for simple and LWR lattice geometries Reviewed International journal

Koji Hayashi, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology Vol. 55 ( 12 ) page： 1434 - 1458 2018.12

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

A prediction method of the true variance of local tally in a Monte Carlo (MC) critical calculation is developed. In the MC calculation, the effective multiplication factor (keff) and the fission rate distribution are estimated by simulating fission chain reactions. The statistical uncertainty of calculation result is commonly estimated by the standard error based on the central limit theorem. However, the evaluated statistical uncertainty would be underestimated when the inter-cycle correlation is not appropriately taken into account. In this study, a theoretical formula of the underestimation ratio (UR) of the statistical uncertainty for a local tally is derived using the eigenfunction expansion and the Autoregressive model. Note that the UR is defined by the ratio of uncertainty estimated by a MC calculation to the true statistical uncertainty. The proposed method is applied to one-dimensional slab and multi-assembly geometries with reflective boundary conditions. In the one-dimensional slab geometry, the prediction results of UR show reasonable agreement with the reference. In the multi-assembly geometry, on the other hand, the prediction results of UR are agreement with the reference regarding the relative spatial shape but there are considerable differences in the absolute values of UR.

DOI： 10.1080/00223131.2018.1513875

Estimation of Sensitivity Coefficient based on Lasso-type Penalized Linear Regression Reviewed International journal

Ryota Katano, Tomohiro Endo, Akio Yamamoto, Kazufumi Tsujimoto

Journal of Nuclear Science and Technology Vol. 55 ( 10 ) page： 1099 - 1109 2018.10

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

We proposed the penalized regression 'adaptive smooth-lasso' for the estimation of sensitivity coefficients of the neutronics parameters. The proposed method utilizes the variation of the microscopic cross-sections and the neutronics parameters obtained by random sampling. The weighted penalty term of the proposed method is more appropriate for the estimation of the sensitivity of neutronics parameters to the microscopic cross-section than that of the conventional methods. In a numerical verification calculation, sensitivity coefficients of keff of an accelerator-driven system are estimated using the proposed method, the conventional penalized regression, and the direct method. Comparison of these results indicates that the proposed method is superior to the conventional penalized linear regression from the viewpoint of reproduction of the reference sensitivity coefficients obtained by the direct method. Through the verification calculations, the proposed method can be a candidate for a practical method to estimate the sensitivity coefficients with low calculation cost.

DOI： 10.1080/00223131.2018.1479988

Web of Science

Uncertainty quantification of criticality in solid-moderated and -reflected cores at Kyoto University Critical Assembly Reviewed International journal

Cheol Ho Pyeon, Masao Yamanaka, Makoto Ito, Go Chiba, Tomohiro Endo, Song Hyun Kim, Willem Fredrik G. van Rooijen

Journal of Nuclear Science and Technology Vol. 55 ( 7 ) page： 812 - 821 2018.7

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

Uncertainty quantification is conducted for the criticality of excess reactivity and control rod worth obtained at the Kyoto University Critical Assembly (KUCA). By combining SRAC2006 and MARBLE code systems, the sensitivity coefficients of the cross sections for aluminum-27 (27Al) comprising mainly of core components are large in the solid-moderated and -reflected cores (A cores) at KUCA. Also, the uncertainty is dominant in the uranium-235 isotope by the covariance data of JENDL-4.0, and a quantitative value is about 150 pcm induced by the JENDL-4.0 data library in the KUCA A cores, whereas the covariance data of 27Al are not prepared in JENDL-4.0. Moreover, the effect of decreasing uncertainty is obtained by applying the cross-sectional adjustment method to the uncertainty analyses. From the results, a series of uncertainty quantifications is expected to clarify the uncertainty of sub-criticality in accelerator-driven system experiments with spallation neutrons in the KUCA A cores.

DOI： 10.1080/00223131.2018.1432426

Scopus

Experimental analysis and uncertainty quantification using random sampling technique for ADS experiments at KUCA Reviewed International journal

Tomohiro Endo, Go Chiba, Willem Frederik Geert van Rooijen, Masao Yamanaka, Cheol Ho Pyeon

Journal of Nuclear Science and Technology Vol. 55 ( 4 ) page： 450 - 459 2018.4

Authorship：Lead author,　Corresponding author Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

Nuclear data-induced uncertainties of neutronics parameters (neutron multiplication factor keff, one-point kinetics parameters and prompt neutron decay constant α) are quantified for lead-bismuth zoned accelerator-driven system experiments at the Kyoto University Critical Assembly, in order to contribute validation for subcritical core analysis. The random sampling technique using SCALE6.2.1/Sampler/NEWT/PARTISN is utilized for the validation and the uncertainty quantification, because the random sampling technique is applicable for a problem which is not easy to apply the perturbation theory. Consequently, it is confirmed that the numerical results of α reasonably agree with the experimental ones, compared with the nuclear data-induced uncertainties. In addition, it is clarified that the nuclear data-induced correlations between α and keff and between α and neutron generation time Λ are strongly negative and positive, respectively. This fact implies that the numerical predictions of keff and Λ can be improved by the data assimilation technique using subcritical experimental results of α, which can be directly measured even for a deep subcritical system.

DOI： 10.1080/00223131.2017.1403387

Scopus

Analysis of the KUCA ADS benchmarks with diffusion theory Reviewed International journal

W.F.G. van Rooijen, T. Endo, G. Chiba, C. H. Pyeon

Progress in Nuclear Energy Vol. 101 page： 243 - 250 2017.11

Language：English Publishing type：Research paper (scientific journal) Publisher：Elsevier B.V.

This manuscript discusses the analysis of the KUCA-A ADS with deterministic theory, specifically diffusion theory. The KUCA-A ADS is a small, thermal-spectrum ADS. For the series of experiments in this manuscript, a DT fusion neutron source was used (14 MeV neutrons). Several sets of KUCA ADS experiments are available as international benchmarks. In our analysis, we calculated the multiplication factor, prompt neutron decay constant (α-eigenvalue), and used time-dependent diffusion theory to analyze pulsed neutron operation of the KUCA-A ADS. The results are that the multiplication factor is generally well predicted as long as the void regions in the core remain of a limited extent. The time-eigenvalue (α-eigenvalue) is also well predicted, and knowledge of the α-mode can be effectively used in the selection of measurement data for curve fitting methods.

DOI： 10.1016/j.pnucene.2017.08.007

Scopus

Benchmarks of subcriticality in accelerator-driven system at Kyoto University Critical Assembly Invited Reviewed International journal

Cheol Ho Pyeon, Masao Yamanaka, Song-Hyun Kim, Thanh-Mai Vu, Tomohiro Endo, Willem Fredrik G. Van Rooijen, Go Chiba

Nuclear Engineering and Technology Vol. 49 ( 6 ) page： 1234 - 1239 2017.9

Language：English Publishing type：Research paper (scientific journal) Publisher：Elsevier B.V.

Basic research on the accelerator-driven system is conducted by combining 235U-fueled and 232Th-loaded cores in the Kyoto University Critical Assembly with the pulsed neutron generator (14 MeV neutrons) and the proton beam accelerator (100 MeV protons with a heavy metal target). The results of experimental subcriticality are presented with a wide range of subcriticality level between near critical and 10,000 pcm, as obtained by the pulsed neutron source method, the Feynman-α method, and the neutron source multiplication method.

DOI： 10.1016/j.net.2017.06.012

Scopus

Recent developments in the GENESIS code based on the Legendre polynomial expansion of angular flux method Invited Reviewed International journal

Akio Yamamoto, Akinori Giho, Tomohiro Endo

Nuclear Engineering and Technology Vol. 49 ( 6 ) page： 1143 - 1156 2017.9

Language：English Publishing type：Research paper (scientific journal) Publisher：Elsevier B.V.

This paper describes recent development activities of the GENESIS code, which is a transport code for heterogeneous three-dimensional geometry, focusing on applications to reactor core analysis. For the treatment of anisotropic scattering, the concept of the simplified Pn method is introduced in order to reduce storage of flux moments. The accuracy of the present method is verified through a benchmark problem. Next, the iteration stability of the GENESIS code for the highly voided condition, which would appear in a severe accident (e.g., design extension) conditions, is discussed. The efficiencies of the coarse mesh finite difference and generalized coarse mesh rebalance acceleration methods are verified with various stabilization techniques. Use of the effective diffusion coefficient and the artificial grid diffusion coefficients are found to be effective to stabilize the acceleration calculation in highly voided conditions.

DOI： 10.1016/j.net.2017.06.016

Web of Science

Scopus

Experimental benchmarks on kinetic parameters in accelerator-driven system with 100 MeV protons at Kyoto University Critical Assembly Reviewed International journal

Cheol Ho Pyeon, Masao Yamanaka, Tomohiro Endo, Willem Fredrik G. van Rooijen, Go Chiba

Annals of Nuclear Energy Vol. 105 page： 346 - 354 2017.7

Language：English Publishing type：Research paper (scientific journal) Publisher：Elsevier B.V.

Accelerator-driven system experiments with spallation neutrons (100 MeV protons and Pb-Bi target) are carried out in the 235U-fueled and Pb-Bi-zoned core at the Kyoto University Critical Assembly, under a subcritical state ranging between 1160 and 11,556 pcm. In these experiments, measurement of the prompt neutron decay constant and the subcriticality is conducted by the pulsed neutron source (PNS) method and the Feynman-α method with the use of optical fiber detectors. The experimental results successfully validate the prompt neutron decay constant and the subcriticality through the deduction of kinetic parameters by both the PNS and the α-fitting methods. The detector position dependency, neutron spectrum and subcriticality measurement methods still remain, however, in these experiments. For onward studies, the experimental benchmarks obtained from these experiments are expected to be involved in the numerical verification of subcriticality on-line monitoring, in the analysis of subcriticality uncertainty and in the deterministic approach to kinetic parameters.

DOI： 10.1016/j.anucene.2017.03.030

Scopus

Estimation of sensitivity coefficients of core characteristics based on reduced-order modeling using sensitivity matrix of assembly characteristics Reviewed International coauthorship International journal

Ryota Katano, Tomohiro Endo, Akio Yamamoto, Mohammad Abdo, Hany Abdel-Khalik

Journal of Nuclear Science and Technology Vol. 54 ( 6 ) page： 637 - 647 2017.6

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

We propose an estimation method of sensitivity coefficients of core neutronics parameters based on a multi-level reduced-order modeling approach. The idea is to use lower-level models to identify the dominant input parameter variations, constrained to the so-called active subspace, which are employed to determine the sensitivity coefficients of the core neutronic parameters. In our implementation, the lower-level model is represented by two-dimensional assembly calculations, which are employed in the preparation of the few-group cross-sections for core-wide calculations. The active subspace basis is estimated using the singular value decomposition of sensitivity matrices of assembly neutronics parameters. In numerical verification calculation, sensitivity coefficients of core characteristics for a typical three-loop PWR equilibrium-cycle are estimated using the proposed method and the direct method. Comparison of these two results shows that the proposed method well reproduces the results obtained by the direct method with lower calculation costs. Through the verification calculations, applicability of the proposed method to practical light water reactor analysis is confirmed.

DOI： 10.1080/00223131.2017.1299052

Scopus

Estimation of modeling approximation errors using data assimilation with the minimum variance approach Reviewed International journal

Akio Yamamoto, Kuniharu Kinoshita, Tomohiro Endo

Journal of Nuclear Science and Technology Vol. 54 ( 4 ) page： 459 - 471 2017.4

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

In this paper, we estimate prediction errors owing to approximations in calculation models (modeling approximation error) using the data assimilation method. Correlations between the modeling approximation error and neutronics parameters obtained through calculations are evaluated in test configurations and then the evaluated correlations are used to predict the modeling approximation errors in design configuration. Formulae to estimate the modeling approximation error using the correlations are derived based on the minimum variance approach and the physical interpretation of the formulae is discussed through simple cases. The proposed method is applied in 2 × 2 and 3 × 3 fuel assembly geometries using specifications of the KAIST benchmark problem. The correlation between the modeling approximation error and parameters (neutron leakage in each fuel assembly) is estimated in 2 × 2 fuel assemblies and then the modeling approximation error in 3 × 3 fuel assemblies is predicted using the correlation. The calculation results not only indicate feasibility of the present method, but also suggest a need for further investigation on the assumptions used in the present study, i.e. applicability and robustness of the correlation among different geometries.

DOI： 10.1080/00223131.2017.1286271

Scopus

GENESIS: A three-dimensional heterogeneous transport solver based on the legendre polynomial expansion of angular flux method Reviewed International journal

Akio Yamamoto, Akinori Giho, Yuki Kato, Tomohiro Endo

Nuclear Science and Engineering Vol. 186 ( 1 ) page： 1 - 22 2017.4

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

A heterogeneous transport solver in three-dimensional (3-D) geometry, GENESIS, is developed incorporating recent developments in the method of characteristics (MOC) in 3-D geometry. The Legendre Polynomial Expansion of Angular Flux (LEAF) method is implemented in the GENESIS code, in which neutron transport is calculated in two-dimensional (2-D) characteristics planes rather than in one-dimensional characteristics lines adopted in the conventional approach of 3-D MOC. Unlike the planar MOC method that combines 2-D MOC calculations through axial leakages, the GENESIS code explicitly considers angular and spatial dependence of outgoing and incoming angular fluxes between axial planes. Thus, the GENESIS code eliminates a crucial approximation used in the planarMOC method: No approximation is used for axial leakage treatment. The GENESIS code can handle flexible shapes of objects in rectangular or hexagonal geometry. A two-level, multigroup generalized coarse mesh rebalance acceleration method is adopted for efficient convergence of neutron transport calculation. Performance of the GENESIS code is verified through various benchmark calculations. The calculation results indicate the fidelity of the GENESIS code based on the LEAF method.

DOI： 10.1080/00295639.2016.1273002

Scopus

Automated generation of burnup chain for reactor analysis applications Reviewed International coauthorship International journal

Viet-Phu Tran, Hoai-Nam Tran, Akio Yamamoto, Tomohiro Endo

Kerntechnik Vol. 82 ( 2 ) page： 196 - 205 2017.4

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

This paper presents the development of an automated generation of burnup chain for reactor analysis applications. Algorithms are proposed to reevaluate decay modes, branching ratios and effective fission product (FP) cumulative yields of a given list of important FPs taking into account intermediate reactions. A new burnup chain is generated using the updated data sources taken from the JENDL FP decay data file 2011 and Fission yields data file 2011. The new burnup chain is output according to the format for the SRAC code system. Verification has been performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Burnup calculations using the new burnup chain have also been performed based on UO2 and MOX fuel pin cells and compared with a reference chain th2cm6fp193bp6T.

DOI： 10.3139/124.110671

Scopus

An efficient execution of Monte Carlo simulation based on delta-tracking method using GPUs Reviewed International journal

Takuya Okubo, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology Vol. 54 ( 1 ) page： 30 - 38 2017.1

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

An efficient execution method for Monte Carlo simulation using graphic processing unit (GPU) is proposed. The delta-tracking method is used since the delta-tracking method can reduce conditional branches and complexity of code implementation, which degrade computational performance on GPUs. In order to improve parallel efficiency in the eigenvalue calculation, generated fission neutrons are recorded using the atomic operation which avoids the data race in GPUs. We also propose a method to efficiently tally neutron flux in a region. The present method is based on the atomic operation and use of fixed-point type number instead of common floating-point type number. The verification calculations using the C5G7 benchmark problem show effectiveness of the proposed numerical algorithms on GPUs through comparison with calculations using central processing units.

DOI： 10.1080/00223131.2016.1202793

Scopus

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Papers - ENDO Tomohiro