論文 - 遠藤　知弘

Underestimation of statistical uncertainty of local tallies in Monte Carlo eigenvalue calculation for simple and LWR lattice geometries 査読有り 国際誌

Koji Hayashi, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology   55 巻 ( 12 ) 頁： 1434 - 1458   2018年12月

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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 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 査読有り 国際誌

Ryota Katano, Tomohiro Endo, Akio Yamamoto, Kazufumi Tsujimoto

Journal of Nuclear Science and Technology   55 巻 ( 10 ) 頁： 1099 - 1109   2018年10月

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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.  

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 査読有り 国際誌

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   55 巻 ( 7 ) 頁： 812 - 821   2018年7月

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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.  

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 査読有り 国際誌

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

Journal of Nuclear Science and Technology   55 巻 ( 4 ) 頁： 450 - 459   2018年4月

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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.  

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

研究データ保存先URL： http://hdl.handle.net/2237/27279

Scopus

Analysis of the KUCA ADS benchmarks with diffusion theory 査読有り 国際誌

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

Progress in Nuclear Energy   101 巻   頁： 243 - 250   2017年11月

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記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：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 招待有り 査読有り 国際誌

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

Nuclear Engineering and Technology   49 巻 ( 6 ) 頁： 1234 - 1239   2017年9月

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記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：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 招待有り 査読有り 国際誌

Akio Yamamoto, Akinori Giho, Tomohiro Endo

Nuclear Engineering and Technology   49 巻 ( 6 ) 頁： 1143 - 1156   2017年9月

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記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：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 査読有り 国際誌

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

Annals of Nuclear Energy   105 巻   頁： 346 - 354   2017年7月

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記述言語：英語   掲載種別：研究論文（学術雑誌）   出版者・発行元：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 査読有り 国際共著 国際誌

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

Journal of Nuclear Science and Technology   54 巻 ( 6 ) 頁： 637 - 647   2017年6月

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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.  

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 査読有り 国際誌

Akio Yamamoto, Kuniharu Kinoshita, Tomohiro Endo

Journal of Nuclear Science and Technology   54 巻 ( 4 ) 頁： 459 - 471   2017年4月

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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 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 査読有り 国際誌

Akio Yamamoto, Akinori Giho, Yuki Kato, Tomohiro Endo

Nuclear Science and Engineering   186 巻 ( 1 ) 頁： 1 - 22   2017年4月

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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 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 査読有り 国際共著 国際誌

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

Kerntechnik   82 巻 ( 2 ) 頁： 196 - 205   2017年4月

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記述言語：英語   掲載種別：研究論文（学術雑誌）  

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 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

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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

Scopus

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

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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

研究データ保存先URL： http://hdl.handle.net/2237/25272

Web of Science

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

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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

Scopus

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

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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

Scopus

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

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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

研究データ保存先URL： http://hdl.handle.net/2237/25270

Scopus

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

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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

Scopus

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

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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

Web of Science

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

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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

Web of Science