研究者詳細 - 遠藤　知弘

論文 - 遠藤　知弘

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An efficient execution of Monte Carlo simulation based on delta-tracking method using GPUs 査読有り国際誌

Takuya Okubo, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 54 巻 ( 1 ) 頁： 30 - 38 2017年1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：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

Reduction of MOC Discretization Errors through a Minimization of Source Ratio Variances 査読有り国際誌

M. Tabuchi, A. Yamamoto, T. Endo, M. Tatsumi

Journal of Nuclear Science and Technology 53 巻 ( 11 ) 頁： 1858 - 1869 2016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：Taylor & Francis, Ltd.

A new technique to reduce discretization errors for ray tracing in the method of characteristics (MOC) is proposed focusing on depletion calculations of single and multi-assembly geometries. In order to efficiently carry out depletion calculations, a calculation scheme using the superhomogenization (SPH) method can be used. However, the discretization errors are caused by changes of neutron sources and total cross sections according to a depletion. This fact means that improvement of accuracy cannot be expected by the calculation scheme with the SPH method when changes of the above parameters are significant. In order to mitigate this problem, a new approach is developed. In the new approach, the discretization errors are reduced by minimizing a variance of a certain parameter which is composed of a ratio of neutron source to total cross section. The verification results suggest that accuracy is degraded by the SPH method as expected especially in a geometry where neutron sources and total cross sections are drastically changing through a depletion. On the other hand, the new approach gives more accurate results compared to the conventional MOC in all calculation cases. Consequently, improvement of calculation efficiency by the new approach is confirmed.

DOI： 10.1080/00223131.2016.1171172

Nuclear data-induced uncertainty quantification of neutronics parameters of accelerator-driven system 査読有り国際誌

Go Chiba, Cheol Ho Pyeon, Wilfred van Rooijen, Tomohiro Endo

Journal of Nuclear Science and Technology 53 巻 ( 10 ) 頁： 1653 - 1661 2016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：Taylor & Francis, Ltd.

Nuclear data-induced uncertainties of neutronics parameters of one accelerator-driven system concept designed by the Japan Atomic Energy Agency are quantified. The variance-covariance data provided in the JENDL-4.0 library are used. Uncertainties are quantified for effective neutron multiplication factor, subcritical neutron multiplication rate, a family of delayed neutron fractions, power peaking and coolant void reactivity at several operational states. Inter-cycle and inter-parameter correlation matrices and detailed information such as nuclide-wise and nuclear data-wise uncertainties are also provided.

DOI： 10.1080/00223131.2015.1127185

A CMFD acceleration method for SP3 advanced nodal method 査読有り国際誌

Akio Yamamoto, Tatsuya Sakamoto, Tomohiro Endo

Nuclear Science and Engineering 184 巻 ( 2 ) 頁： 168 - 173 2016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：Taylor & Francis, Ltd.

Flux-level-fixup (FF) coarse-mesh finite difference (CMFD) (FF-CMFD), which increases numerical stability during nonlinear iterations for the SP3 advanced nodal method, is proposed as an improved CMFD implementation. In contrast to the scalar flux that appeared in the advanced nodal method with diffusion theory, the second flux moment φ2 in the SP3 method could take a very small value since it can take both positive and negative values in a node. This is a root cause of inefficiency of the SP3 advanced nodal method when conventional CMFD acceleration is directly applied. In the proposed FF-CMFD method, a constant value is added to the second flux moment φ2 to fix up its value to a sufficiently large positive value for stable numerical calculations. The efficiency of the FF-CMFD method is verified through benchmark calculations.

DOI： 10.13182/NSE16-53

Statistical Error Estimation of the Feynman-α Method using the Bootstrap Method 査読有り国際誌

Tomohiro Endo, Akio Yamamoto, Takahiro Yagi, Cheol Ho Pyeon

Journal of Nuclear Science and Technology 53 巻 ( 9 ) 頁： 1447 - 1453 2016年9月

担当区分：筆頭著者,　責任著者記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis, Ltd.

Applicability of the bootstrap method is investigated to estimate the statistical error of the Feynman-alpha method, which is one of the subcritical measurement techniques on the basis of reactor noise analysis. In the Feynman-alpha method, the statistical error can be simply estimated from multiple measurements of reactor noise, however it requires additional measurement time to repeat the multiple times of measurements. Using a resampling technique called "bootstrap method," standard deviation and confidence interval of measurement results obtained by the Feynman-α method can be estimated as the statistical error, using only a single measurement of reactor noise. In order to validate our proposed technique, we carried out a passive measurement of reactor noise without any external source, i.e. with only inherent neutron source by spontaneous fission and (alpha,n) reactions in nuclear fuels at the Kyoto University Criticality Assembly. Through the actual measurement, it is confirmed that the bootstrap method is applicable to approximately estimate the statistical error of measurement results obtained by the Feynman-alpha method.

DOI： 10.1080/00223131.2015.1113898

Discontinuity factors for simplified P3 theory 査読有り国際誌

Akio Yamamoto, Tatsuya Sakamoto, Tomohiro Endo

Nuclear Science and Engineering 183 巻 ( 1 ) 頁： 39 - 51 2016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：American Nuclear Society

New discontinuity factors (DFs), i.e., individual and common DFs, for the simplified P3 (SP3) theory are proposed. In the individual DFs, two DFs are used for zeroth- and second-order angular moments in order to preserve first- and third-order angular moments of SP3 at a surface of the homogenized region. Contrarily, the same value of DF is used for zeroth- and second-order angular moments, and the first-order angular moment is preserved in the common DF. Theoretical derivation for these DFs are described, and then, actual numerical calculation procedures for these DFs are explained. Verification results in color-set geometries loaded with UO2 and mixed oxide fuel assemblies indicate the validity of the present method for cell-homogenized pin-by-pin SP3 calculations. Homogenization errors on keff and pin-power distribution are significantly reduced by the present DFs. The proposed DFs can be used for practical pin-by-pin core analyses using the SP3 theory.

DOI： 10.13182/NSE15-102

Uncertainty Quantification of LWR Core Characteristics Using Random Sampling Method 査読有り国際誌

Akio Yamamoto, Kuniharu Kinoshita, Tomoaki Watanabe, Tomohiro Endo, Yasuhiro Kodama, Yasunori Ohoka, Tadashi Ushio, Hiroaki Nagano

Nuclear Science and Engineering 181 巻 ( 2 ) 頁： 160 - 174 2015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：Taylor & Francis, Ltd.

Uncertainties of various neutronics characteristics in commercial boiling water reactor (BWR) and pressurized water reactor (PWR) cores due to cross-section covariance are evaluated by the Latin Hypercube Sampling (LHS) method, which is an efficient random sampling algorithm. Thermal-hydraulic feedback and burnup effects are fully and explicitly taken into account using a licensing-grade core simulator. Uncertainties for various core characteristics are evaluated by the statistical processing of core calculation results based on the LHS method. The calculation results indicate that uncertainty of critical eigenvalue (i.e., core reactivity) in the BWR core is comparable to that of a typical PWR core. On the other hand, uncertainties of assembly relative power distribution and maximum assembly burnup in the present BWR core are much smaller than those of the present PWR core. The strong thermal-hydraulic feedback effect in the BWR core significantly contributes to the difference of uncertainties in BWR and PWR cores.

DOI： 10.13182/NSE14-152

Application of correction technique using leakage index combined with SPH or discontinuity factors for energy collapsing on pin-by-pin BWR core analysis 査読有り国際誌

Tatsuya Fujita, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 52 巻 ( 3 ) 頁： 355 - 370 2015年3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：Taylor & Francis, Ltd.

A correction technique to capture the spectral interference effect on collapsed cross sections is combined with the superhomogenization (SPH) factor or the discontinuity factor (DF) and is applied to the pin-by-pin core analysis for boiling water reactors (BWRs). The spectral interference effect has relationship with variations of neutron leakage in each pin-cell from the viewpoint of neutron balance. In order to correct collapsed cross sections, a new correction technique, in which the neutron leakage in each pin-cell is used as a correction index, was proposed in the previous study. By this correction technique, the reference coarse group cross sections are well reproduced and the calculation accuracies are improved. However, the reference fine group calculation results could not be reproduced since the correction technique cannot reduce energy collapsing errors. Thus, we combine the correction technique with the SPH factor or the DF to reduce energy collapsing errors. In order to verify and discuss the applicability of the correction technique with the SPH factor or the DF, two-dimensional benchmark calculations considering typical characteristics of BWR cores are carried out. The correction technique with the DF more accurately reproduces the reference fine group calculation results than that with the SPH factor.

DOI： 10.1080/00223131.2014.948520

Applicability of angular flux discontinuity factor preserving region-wise leakage for integro-differential transport equation 査読有り国際誌

Tatsuya Sakamoto, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 51 巻 ( 10 ) 頁： 1264 - 1273 2014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：Taylor & Francis, Ltd.

In the current core analysis, spatial homogenization is utilized to reduce the computational time. The discontinuity factor (DF) is one of the effective correction factors to reduce spatial homogenization error. The DF in diffusion equation is widely used on the other hand the DF in transport equation has not been put to practical use although several efforts have been carried out. In this paper, the angular flux discontinuity factor (AFDF) as the DF for the integro-differential transport equation (e.g., the discrete-ordinate method, the method of characteristics) is theoretically described and its applicability is discussed. The AFDF is used to preserve the region-wise neutron leakage at each spatial mesh and defined as a ratio of heterogeneous and homogeneous angular fluxes at the homogenized region surface. In a homogeneous calculation with the AFDF, the angular flux is discontinuous at the region surface. In this paper the applicability of the AFDF to fuel pin cell homogenization is verified for one-dimensional slab geometry. As a result of this verification, it is confirmed that the AFDF has the capability to reduce the spatial homogenization error of fuel pin cell homogenization.

DOI： 10.1080/00223131.2014.919882

A new technique for spectral interference correction on pin-by-pin BWR core analysis 査読有り国際誌

Tatsuya Fujita, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 51 巻 ( 6 ) 頁： 783 - 797 2014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：Taylor & Francis, Ltd.

A new correction technique to capture the spectral interference effect on collapsed cross sections, which focuses on application to the pin-by-pin boiling water reactor (BWR) core analysis, is proposed. The spectral interference effect, which is caused by adjacent loadings of different types of fuel assemblies, has relationship with variations of neutron leakage in each pin-cell from the viewpoint of neutron balance. Variation of neutron leakage affects neutron spectrum and thus the neutron leakage is considered to be important to correct coarse-group cross sections used in core calculations. We focus on the neutron leakage in each pin-cell and use it as a correction index (i.e., a leakage index (LI)), which is defined as the volume-averaged neutron leakage in a pin-cell. By utilizing the leakage index, we represent the variations of coarse-group cross sections as the linear combination of LIs. In order to verify and discuss the applicability of the present correction technique, two-dimensional benchmark calculations considering typical characteristics of BWR cores are carried out. From the calculation results, the present correction technique well reproduces the reference coarse-group cross sections and improves the calculation accuracies.

DOI： 10.1080/00223131.2014.903212

Cross section adjustment method based on random sampling technique 査読有り国際誌

Tomoaki Watanabe, Tomohiro Endo, Akio Yamamoto, Yasuhiro Kodama, Yasunori Ohoka, Tadashi Ushio

Journal of Nuclear Science and Technology 51 巻 ( 5 ) 頁： 590 - 599 2014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：Taylor & Francis, Ltd.

A cross section adjustment method based on the random sampling technique is proposed. In the proposed method, correlations among cross sections and core parameters are used instead of sensitivity coefficients of cross sections, which are necessary in the conventional method. The correlations are statistically estimated by the random sampling technique. The proposed method is theoretically consistent with the conventional method and provides comparable adjusted cross sections when sufficient number of random sampling is taken into account. The proposed method would be suitable for practical light water reactor (LWR) core analysis since estimation of sensitivity coefficients, which requires considerable computational cost in typical LWR problems, is not necessary. Through a benchmark problem in simple pin-cell geometry, adjusted cross sections by the present and the conventional cross section adjustment method are compared. The adjusted cross sections by the present method well reproduce the conventional ones, thus the feasibility of the present method is confirmed.

DOI： 10.1080/00223131.2014.882801

A macroscopic cross-section model for BWR pin-by-pin core analysis 査読有り国際誌

Tatsuya Fujita, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 51 巻 ( 3 ) 頁： 282 - 304 2014年3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：Taylor & Francis, Ltd.

A macroscopic cross-section model used in boiling water reactor (BWR) pin-by-pin core analysis is studied. In the pin-by-pin core calculation method, pin-cell averaged cross sections are calculated for many combinations of core state and depletion history variables and are tabulated prior to core calculations. Variations of cross sections in a core simulator are caused by two different phenomena (i.e. instantaneous and history effects). We treat them through the core state variables and the exposure-averaged core state variables, respectively. Furthermore, the cross-term effect among the core state and the depletion history variables is considered. In order to confirm the calculation accuracy and discuss the treatment of the cross-term effect, the k-infinity and the pin-by-pin fission rate distributions in a single fuel assembly geometry are compared. Some cross-term effects could be negligible since the impacts of them are sufficiently small. However, the cross-term effects among the control rod history (or the void history) and other variables have large impacts
thus, the consideration of them is crucial. The present macroscopic cross-section model, which considers such dominant cross-term effects, well reproduces the reference results and can be a candidate in practical applications for BWR pin-by-pin core analysis on the normal operations.

DOI： 10.1080/00223131.2014.864248

Application of augmented reality to nuclear reactor core simulation for fundamental nuclear engineering education 査読有り国際誌

Kosuke Tsujita, Tomohiro Endo, Akio Yamamoto

Nuclear Technology 185 巻 ( 1 ) 頁： 71 - 84 2014年1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：Taylor & Francis, Ltd.

A nuclear reactor core simulation system using augmented reality (AR) has been developed. Augmented reality is a technology that can provide additional information by overlaying computer graphics onto the image of actual world. In the past, AR has been applied to operation assistance in various fields. In the field of nuclear engineering, AR has been applied to support the decommissioning of nuclear power plants. Using AR, workers can simultaneously see the image of the actual world and helpful information
thus, they can intuitively imagine their works. This advantage of AR can be applied not only to operation assistance but also to other purposes. Therefore in this study we have tried to apply AR to a nuclear core simulator. The major purpose of the present AR core simulator is education of novice students who are not very familiar with nuclear reactor core behavior by enabling direct "hand manipulation" of a reactor model in the real world. For example users can directly insertlwithdraw a control rod and can directly see the power variation of the reactor through the AR technology.

DOI： 10.13182/NT13-7

Uncertainty estimation of core safety parameters using cross-correlations of covariance matrix 査読有り国際誌

Akio Yamamoto, Yoshihiro Yasue, Tomohiro Endo, Yasuhiro Kodama, Yasunori Ohoka, Masahiro Tatsumi

Journal of Nuclear Science and Technology 50 巻 ( 10 ) 頁： 966 - 978 2013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：Taylor & Francis, Ltd.

An uncertainty reduction method for core safety parameters, for which measurement values are not obtained, is proposed. We empirically recognize that there exist some correlations among the prediction errors of core safety parameters, e.g., a correlation between the control rod worth and the assembly relative power at corresponding position. Correlations of errors among core safety parameters are theoretically estimated using the covariance of cross sections and sensitivity coefficients of core parameters. The estimated correlations of errors among core safety parameters are verified through the direct Monte Carlo sampling method. Once the correlation of errors among core safety parameters is known, we can estimate the uncertainty of a safety parameter for which measurement value is not obtained.

DOI： 10.1080/00223131.2013.820155

Preservation of transmission probabilities in the method of characteristics 査読有り国際誌

Masato Tabuchi, Akio Yamamoto, Tomohiro Endo, Naoki Sugimura

Journal of Nuclear Science and Technology 50 巻 ( 8 ) 頁： 837 - 843 2013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：Taylor & Francis, Ltd.

An efficient calculation scheme for the method of characteristics, which allows coarser ray separation width, is newly proposed. Effective lengths of characteristics lines are evaluated from the viewpoint of preservation of transmission probability in the present method. The effective lengths to preserve transmission probabilities can be rigorously estimated through detailed numerical integrations. However, calculation and storage of all effective lengths is inefficient since the transmission probabilities depend on many parameters such as azimuthal angle, polar angle, position of characteristics line, and total cross section. To resolve this issue, in this paper, the analytical equations for the effective lengths are derived from an assumption that distribution function of actual characteristics lengths can be expressed by a linear function. The new approach introduced in this paper is named as transmission probability preservation through linear approximation (TPPL). Verification results indicate that accurate results can be obtained by the TPPL even if ray separation width becomes coarse.

DOI： 10.1080/00223131.2013.807753

Convergence analysis of MOC inner iterations with large negative self-scattering cross-section 査読有り国際誌

Masato Tabuchi, Akio Yamamoto, Tomohiro Endo, Naoki Sugimura

Journal of Nuclear Science and Technology 50 巻 ( 5 ) 頁： 493 - 502 2013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：Taylor & Francis, Ltd.

When the transport correction is applied to the total cross-section, the self-scattering cross-section could have a negative value in order to preserve the balance of partial cross-sections. The negative self-scattering cross-section may lead to a negative impact on the convergence behavior for the method of characteristics (MOC), especially in a problem with large moderator regions containing hydrogen. In order to address this issue, the spectral radius of the inner iteration of MOC is theoretically estimated for various self-scattering cross-sections. It is found that the spectral radius of the inner iteration of MOC could exceed unity for a large negative self-scattering cross-section, which results in numerical divergence. A countermeasure for the divergence using the successive over relaxation method is also discussed in this paper.

DOI： 10.1080/00223131.2013.785271

Analysis of integral experiment on erbia-loaded thermal spectrum cores using Kyoto University Critical Assembly by MCNP code with various cross section libraries 査読有り国際誌

Yevgeniy Tur, Tomohiro Endo, Akio Yamamoto, Hironobu Unesaki, Masatoshi Yamasaki

Journal of Nuclear Science and Technology 49 巻 ( 10 ) 頁： 1028 - 1041 2012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：Taylor & Francis, Ltd.

Under the project on high burnup nuclear fuel development using erbium as a burnable poison, a series of experiments were performed at the Kyoto University Critical Assembly. The experimental results have formed the basis for this study which aims to analyze the suitability of various evaluated nuclear data libraries for using them in neutronic calculations under the project. The MCNP code was used for the analysis. Calculation model geometry was fully detailed, and ENDF, JENDL, JEFF, and TENDL libraries were used during calculation. For the cross sections of erbium nuclides, the analysis revealed that calculated results upon all the libraries corresponded with experimental data within the errors. However, in some libraries, significant differences were found in case of carbon and uranium nuclides under certain conditions.

DOI： 10.1080/00223131.2012.723174

An optimization approach to establish an appropriate energy group structure for BWR pin-by-pin core analysis 査読有り国際誌

Tatsuya Fujita, Kenichi Tada, Tomohiro Endo, Akio Yamamoto, Shinya Kosaka, Go Hirano, Kenichiro Nozaki

Journal of Nuclear Science and Technology 49 巻 ( 7 ) 頁： 689 - 707 2012年7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：Taylor & Francis, Ltd.

An optimization approach to establish an appropriate multi-group energy structure for boiling water reactor (BWR) pin-by-pin finemesh core analysis is proposed. In the present approach, the number of energy groups of cross sections is successively reduced or increased. In order to select an energy group boundary that is removed or added, performances of all possible candidates of energy group structures are tested in multi-assembly geometries. Then, the energy group boundary, which provides theminimumdifference of the k-infinity or the pin-by-pin fission rate distribution, is finally removed or added. This procedure is repeated until the number of energy groups reaches to the target value. In order to confirm the applicability of the present approach, the accuracies of the k-infinity and the pin-by-pin fission rate distribution are investigated in various 2 × 2 multi-assembly geometries with the established energy group structure. From the verification results, the differences of the k-infinity and the pin-by-pin fission rate distribution between the reference (fine) and the established (coarse) energy group structure are small in the various 2 × 2 multiassembly geometries. Therefore, we can conclude that the present approach is efficient to establish an appropriate energy group structure for BWR pin-by-pin fine mesh core analysis.

DOI： 10.1080/00223131.2012.693890

A unified approach for numerical calculation of space-dependent kinetic equation 査読有り国際誌

Yuichiro Ban, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 49 巻 ( 5 ) 頁： 496 - 515 2012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：Taylor & Francis, Ltd.

A unified numerical method based on the factorization approach is developed to solve the spacedependent neutron kinetic equation. Various numerical methods for solving the space-dependent kinetic equation have been developed so far. These methods can be classified into two categories, i.e., the direct and the factorization methods. The factorization method is known as an effective numerical method. In the present study, a new factorization method named the multigrid amplitude function (MAF) method is developed. Unlike the improved quasi-static (IQS) method, an independent amplitude function is assigned for each spatial region and energy group in the MAF method. The MAF method is a generalization of conventional methods, e.g., the frequency transform, IQS, and Theta methods. To evaluate the amplitude function in the MAF method, the time-dependent coarse-mesh finite difference (TCMFD) method is developed. The MAF method is implemented into a space-dependent kinetic code on the basis of the analytical polynomial nodal method. In order to verify the effectiveness of the MAF method, the TWIGL, Langenbuch, Maurer, and Werner (LMW), and Laboratorium fü r Reaktorregelung and Anlagensicherung (LRA) benchmark problems are analyzed. The calculation results show the effectiveness of the present method.

DOI： 10.1080/00223131.2012.677126

Application of quick subchannel analysis method for three-dimensional pin-by-pin BWR core calculations 査読有り国際誌

Kenichi Tada, Tatsuya Fujita, Tomohiro Endo, Akio Yamamoto, Shinya Kosaka, Go Hirano, Kenichiro Nozaki

Journal of Nuclear Science and Technology 48 巻 ( 12 ) 頁： 1437 - 1452 2011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：Taylor & Francis, Ltd.

Three-dimensional pin-by-pin core analysis is considered to be a candidate for the next-generation BWR core calculation method. In our previous study, the applicability of the transport and burnup calculations for a three-dimensional pin-by-pin BWR core analysis was investigated. However, the thermal-hydraulics calculation has not yet been studied in this framework. In the conventional core analysis code, the bundlewise thermal-hydraulics calculation is adopted. In the actual core analysis, the power distribution inside a fuel assembly is tilted at the region adjacent to a control blade or the core peripheral region. In these regions, the consideration of the subchannel-wise void distribution has an impact on the fission rate distribution. Therefore, an evaluation of the detailed void distribution inside an assembly, i.e., the incorporation of the subchannel wise void distribution, is desirable for the pin-by-pin BWR core analysis. Although several subchannel analysis codes have been developed, these subchannel analysis codes generally require a large computational effort to estimate the subchannel-wise void distribution in a whole BWR core. Therefore, to analyze a whole BWR core within a reasonable computation time, it was necessary to apply a fast subchannel analysis code. In this paper, a quick subchannel analysis code dedicated to pin-by-pin BWR core analysis is newly developed, and the void distribution of the present subchannel analysis code is compared with the prevailing subchannel analysis code NASCA using three-dimensional single-assembly geometries. Since the present subchannel analysis code is used for a coupled neutronics/thermal-hydraulics analysis, the results of the coupling calculation are also compared with those of NASCA. The calculation result indicates that the void distribution difference between NASCA and the present subchannel analysis code is slightly less than 10%. This result indicates that the prediction accuracy of the present subchannel analysis code will be reasonably appropriate for a pin-by-pin BWR core analysis. Furthermore, the results show that the calculation time of the present subchannel analysis code is only 10 min for a hypothetical three-dimensional ABWR quarter-core geometry using a single CPU. This calculation time is sufficient for a pin-by-pin BWR core analysis.

DOI： 10.1080/18811248.2011.9711837

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