研究者詳細 - 遠藤　知弘

MISC - 遠藤　知弘

分割表示 188 件中 141 - 160 件目／全件表示 >>

Impact of Nearest Neighbor Distribution of Fuel Particle on Neutronics Characteristics in Statistical Geometry Model 査読有り国際誌

T. Koide, T. Endo, A. Yamamoto, K. Kirimura, K. Yamaji

Proc. PHYSOR2014CD-ROM 巻頁： 1101295 2014年9月

Generation of Simplified Burnup Chain using Contribution Matrix of Nuclide Production 査読有り国際誌

R. Katano, A. Yamamoto, T. Endo, Y. Kamiyama, K. Kirimura, S. Kosaka

Proc. PHYSOR2014CD-ROM 巻頁： 1101303 2014年9月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：Japan Atomic Energy Agency

A procedure to automatically generate a simplified burnup chain is proposed, which preserves prediction accuracy of target nuclides number density using calculation results of a detail burnup chain. In the present procedure, the contributions to production of the target nuclides are quantitatively evaluated at first using the off-diagonal elements of a burnup matrix and nuclides number density. Then, by considering the contributions to nuclide production, which is represented as a “contribution matrix”, the nuclides in detail burnup chain are selected according to the importance represented as the contribution matrix. Finally, a simplified burnup chain is generated with the target nuclides and the selected nuclides from the contribution matrix. Since the present method only utilizes the burnup matrix, which is usually used in the common lattice physics computation, its implementation to production codes is easy. In order to examine the effectiveness of the present method, burnup calculation in a PWR fuel pin-cell problem with some simplification on calculation conditions is carried out. The calculation results suggest that the present method will be a good candidate for an automated generation method for a simplified burnup chain.

DOI： 10.11484/jaea-conf-2014-003

Uncertainty Quantification of BWR Core Characteristics using Latin Hypercube Sampling Method 査読有り国際誌

K. Kinoshita, A. Yamamoto, T. Endo, Y. Kodama, Y. Ohoka, T. Ushio, H. Nagano

Proc. PHYSOR2014CD-ROM 巻頁： 1090063 2014年9月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：Japan Atomic Energy Agency

Uncertainties of neutronics characteristics in a commercial BWR core due to cross section covariance are evaluated by the Latin Hypercube Sampling (LHS) method, which is a Monte-Carlo based efficient sampling algorithms. The Peach Bottom Unit 2, which is a BWR core, is a target for the uncertainty quantification in the present study. Thermal hydraulics feedback and burnup effects are fully and explicitly taken into account using a licensing-grade core simulator. Uncertainties for various core characteristics and correlations among them 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 typical PWR. 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 typical PWR. The strong thermal hydraulics feedback effect of BWR would significantly contribute the difference of uncertainties in BWR and PWR.

DOI： 10.11484/jaea-conf-2014-003

Uncertainty Quantification of Neutronics Characteristics using Latin Hypercube Sampling Method 査読有り国際誌

K. Kinoshita, A. Yamamoto, T. Endo, Y. Kodama, Y. Ohoka, T. Ushio, H. Nagano

Proc. PHYSOR2014CD-ROM 巻頁： 1101192 2014年9月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：Japan Atomic Energy Agency

The Latin Hypercube Sampling (LHS) method is applied to the uncertainty quantification of neutronics characteristics based on the Monte-Carlo based sampling method. The Monte-Carlo based sampling method is one of the uncertainty quantification methods of neutronics characteristics (e.g. neutron multiplication factor) due to uncertainty cross section. The sampling method has the advantage that burnup and thermal-hydraulics feedback effects are easily considered in complicated light water reactor core analysis. Contrary, the sampling method has the disadvantage that statistical errors are involved in estimated uncertainties of neutronics characteristics since the sampling method is a probabilistic approach. Although the statistical errors can be reduced by increasing the number of samples (e.g. number of perturbed cross section libraries), computational cost also increases. Therefore development of an efficient uncertainty quantification method, which provides smaller statistical error of estimated uncertainty while reducing number of samples, is highly desirable. In the present study, adequacy and performance of the LHS method is investigated as an efficient Monte-Carlo based sampling method. Uncertainty quantification of multiplication factor for a BWR fuel assembly is carried out with the LHS method. The results indicate that uncertainty quantification using the LHS method is more efficient than that using the conventional sampling method, which utilizes simple random sampling of cross sections.

DOI： 10.11484/jaea-conf-2014-003

Theoretical Prediction on Underestimation of Statistical Uncertainty for Fission Rate Tally in Monte Carlo Calculation 査読有り国際誌

T. Endo, A. Yamamoto, K. Sakata

Proc. PHYSOR2014CD-ROM 巻頁： 1126624 2014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：研究発表ペーパー・要旨（国際会議）出版者・発行元：Japan Atomic Energy Agency

A theoretical model to predict underestimation of standard deviation for the mean of fission rate in each tally region is proposed on the basis of the behavior of higher order modes in fission source distribution and the autoregressive (AR) model. The predicted underestimation of standard deviation for the mean of fission rate is compared with that obtained by actual Monte Carlo calculations and they agree well each other. Dependency of underestimation of standard deviation on number of tally regions and spatial position is clarified by the proposed theoretical model. The present theoretical model can be used to quantitatively predict underestimation of standard deviation of local fission rate tally.

DOI： 10.11484/jaea-conf-2014-003

Applicability of the Cross Section Adjustment Method Based on Random Sampling Technique for Burnup Calculation 査読有り国際誌

T. Watanabe, T. Endo, A. Yamamoto, Y. Kodama, Y. Ohoka, T. Ushio

Proc. PHYSOR2014CD-ROM 巻頁： 1098930 2014年9月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：Japan Atomic Energy Agency

Applicability of the cross section adjustment method based on random sampling (RS) technique to burnup calculations is investigated. The cross section adjustment method is a technique for reduction of prediction uncertainties in reactor core analysis and has been widely applied to fast reactors. As a practical method, the cross section adjustment method based on RS technique is newly developed for application to light water reactors (LWRs). In this method, covariance among cross sections and neutronics parameters are statistically estimated by the RS technique and cross sections are adjusted without calculation of sensitivity coefficients of neutronics parameters, which are necessary in the conventional cross section adjustment method. Since sensitivity coefficients are not used, the RS-based method is expected to be practically applied to LWR core analysis, in which considerable computational costs are required for estimation of sensitivity coefficients. Through a simple pin-cell burnup calculation, applicability of the present method to burnup calculations is investigated. The calculation results indicate that the present method can adequately adjust cross sections including burnup characteristics.

DOI： 10.11484/jaea-conf-2014-003

Prediction on Underestimation of Variance for Fission Rate Distribution in Monte-Carlo Calculation 査読有り国際誌

A. Yamamoto, K. Sakata, T. Endo

Trans. Am. Nucl. Soc.110 巻頁： 515 - 518 2014年6月

An Improved CMFD Acceleration for SP3 Advanced Nodal Method 査読有り国際誌

T. Sakamoto, A. Yamamoto, T. Endo

Trans. Am. Nucl. Soc.110 巻頁： 535 - 537 2014年6月

Evaluation of Higher Mode Components of Fission Source Distribution in Monte-Carlo Calculation 査読有り国際誌

K. Sakata, T. Endo, A. Yamamoto

Proc. SNA-MC 2013 頁： 03503 2014年6月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）出版者・発行元：EDP Sciences

In order to accurately estimate uncertainty of fission source distribution in a Monte Carlo calculation, a theoretical model predicting behavior of higher order modes in fission source distribution at each cycle is proposed considering the inter-cycle correlation. In a Monte Carlo calculation, statistical noise stimulates higher order modes in fission source distribution at each cycle thus the fission source distribution may not be fully converged to a fundamental mode even if number of power iterations is sufficiently large. This is especially true for a large geometry with high dominance ratio and is one of the root cause of underestimation of statistical uncertainty. In the present model, a fission source distribution is expanded by a series of higher order modes and their behaviors are estimated by variation of expansion coefficients at each cycle. The expansion coefficients show fluctuation due to statistical noise during a Monte Carlo calculation. By observing decay ratio of higher order modes in PWR single assembly geometry, validity of the present theoretical model is confirmed. Furthermore, behaviors of higher order mode components in PWR single or multi fuel assemblies geometry are investigated.

DOI： 10.1051/snamc/201403503

Estimation of Self-Shielding Effect on Uncertainty of Neutronics Characteristics using Random Sampling Method and Continuous-Energy Slowing-Down Calculation 査読有り国際誌

A. Yamamoto, S. Sato, T. Endo

Trans. Am. Nucl. Soc.109 巻頁： 1436 - 1438 2013年11月

Subcriticality Measurement Technique using Inherent Neutron Source in Uranium Fuel 査読有り国際誌

T. Shiozawa, T. Endo, A. Yamamoto, C.H. Pyeon, T. Yagi

Trans. Am. Nucl. Soc.109 巻頁： 826 - 829 2013年11月

Behavior of Higher Order Fission Source Distribution in Monte-Carlo Calculations 査読有り国際誌

A. Yamamoto, K. Sakata, T. Endo

Trans. Am. Nucl. Soc.109 巻頁： 1361 - 1364 2013年11月

Uncertainty and Correlation Estimation of Reload Safety Parameters of PWR using Random Sampling Method 査読有り国際誌

T. Watanabe, T. Endo, A. Yamamoto, Y. Kodama, Y. Ohoka, T. Ushio

Trans. Am. Nucl. Soc.109 巻頁： 1365 - 1368 2013年11月

Few-Group Macroscopic Cross Section Adjustment for LWRs using Random Sampling Technique 査読有り国際誌

A. Yamamoto, S. Kato, T. Endo

Trans. Am. Nucl. Soc.108 巻頁： 894 - 897 2013年6月

Study on Discontinuity Factor for Angular Flux in Transport Equation 査読有り国際誌

T. Sakamoto, T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc.108 巻頁： 887 - 890 2013年6月

Reduction of Discretization Error for Ray Tracing of MOC Through a Correction on Collision Probabilities 査読有り国際誌

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

Proc. M&C2013 頁： 1853 - 1864 2013年5月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）

A new correction model for ray tracing of the method of characteristics is proposed in order to reduce discretization error. As the ray tracing parameters such as azimuthal angle division, polar angle division and ray separation are considered in this study. In the method of characteristics, region average scalar fluxes can be implicitly expressed by collision probabilities, although these collision probabilities are not directly treated in the ordinary calculation scheme. From this viewpoint, difference between a coarse ray tracing condition and a detailed one can be interpreted as the difference in the estimation of collision probabilities. In other words, the discretization error for ray tracing can be recognized as a consequence of inaccurate collision probabilities caused by coarse ray tracing. This discussion suggests that accurate region average scalar flux can be obtained through an appropriate correction on collision probabilities. In this paper, a correction model on collision probabilities is theoretically derived based on the neutron balance equation, and its validity is confirmed through typical single assembly calculations. The effectiveness of the present correction method is also discussed in this paper. It is confirmed that discretization error for ray tracing can be significantly reduced by the present correction method in a multi-assembly calculation, though the correction factor is estimated in single assembly geometry.

Explicit Estimation of Higher Order Modes in Fission Source Distribution of Monte-Carlo Calculation 査読有り国際誌

A. Yamamoto, K. Sakata, T. Endo

Proc. M&C2013 頁： 2814 - 2827 2013年5月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）

Magnitude of higher order modes in fission source distribution of a multi-group Monte-Carlo calculation is estimated using the orthogonal property of forward and adjoint fission source distributions. Calculation capability of the forward and adjoint fission source distributions for fundamental and higher order modes are implemented in the AEGIS code, which is a twodimensional transport code based on the method of characteristics. With the calculation results of the AEGIS code, magnitudes of the first to fifth higher order modes in fission source distribution obtained by the multi-group Monte-Carlo code GMVP are estimated. There are two contributions in the present study - (1) establishment of a surrogate model, which represents convergence of fission source distribution taking into account the inherent statistical "noise" of higher order modes of Monte-Carlo calculations and (2) independent confirmation of the estimated dominance ratio in a Monte-Carlo calculation. The surrogate model would contribute to studies of the intercycle correlation and estimation of sufficient number of inactive/active cycles.

Higher Order Treatment on Temporal Derivative of Angular Flux for Time-Dependent MOC 査読有り国際誌

K. Tsujita, T. Endo, A. Yamamoto, Y. Kamiyama, K. Kirimura

Proc. M&C2013 頁： 1826 - 1838 2013年5月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）

A new kinetic analysis method, whose angular dependence of temporal derivative for angular flux is accurately treated within practical memory requirement, is proposed. The method of characteristics (MOC) is being widely used for reactor analysis thanks to the advances of numerical algorithms and computer hardware. However, the computational resources, i.e., the memory capacity, can be still a crucial problem for rigorous kinetic calculations using MOC. In the straightforward approach for kinetic calculation using MOC, the segment-averaged angular fluxes should be stored on the memory in order to explicitly calculate the temporal derivative of the angular flux, which would require huge memory. Thus, in the conventional kinetic calculation code using MOC, the temporal derivative of the angular flux has been approximated as angularly isotropic in order to reduce the memory requirement (isotropic assumption). However, the approximation error caused by the conventional isotropic assumption has not been thoroughly and quantitatively investigated so far and an accurate kinetic calculation method, which can quantitatively estimate the above approximation error within practical memory storage, has not been developed. The present study tries to address this issue with a newly developed approach. Effect of the approximate treatment for the temporal derivative of angular flux is evaluated through benchmark calculations.

Application of the Multigrid Amplitude Function Method for Time-Dependent Transport Equation using MOC 査読有り国際誌

K. Tsujita, T. Endo, A. Yamamoto

Proc. M&C2013 頁： 1839 - 1852 2013年5月

記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）

An efficient numerical method for time-dependent transport equation, the mutigrid amplitude function (MAF) method, is proposed. The method of characteristics (MOC) is being widely used for reactor analysis thanks to the advances of numerical algorithms and computer hardware. However, efficient kinetic calculation method for MOC is still desirable since it requires significant computation time. Various efficient numerical methods for solving the space-dependent kinetic equation, e.g., the improved quasi-static (IQS) and the frequency transform methods, have been developed so far mainly for diffusion calculation. These calculation methods are known as effective numerical methods and they offer a way for faster computation. However, they have not been applied to the kinetic calculation method using MOC as the authors’ knowledge. Thus, the MAF method is applied to the kinetic calculation using MOC aiming to reduce computation time. The MAF method is a unified numerical framework of conventional kinetic calculation methods, e.g., the IQS, the frequency transform, and the theta methods. Although the MAF method is originally developed for the space-dependent kinetic calculation based on the diffusion theory, it is extended to transport theory in the present study. The accuracy and computational time are evaluated though the TWIGL benchmark problem. The calculation results show the effectiveness of the MAF method.

Random Sampling-Based Cross-Section Adjustment Technique for LWR Core Analysis 査読有り国際誌

S. Kato, T. Endo, A. Yamamoto, H. Yamauchi, Y. Kimura

Proc. Int. Congress on Advances in Nuclear Power Plants (ICAPP2013) 頁： 877 - 883 2013年4月

担当区分：責任著者記述言語：英語掲載種別：研究発表ペーパー・要旨（国際会議）

In the present study, a random sampling-based cross-section adjustment technique is developed to apply to LWR reactor core analysis procedure. In this technique, random sampling method is utilized for fine energy group microscopic cross-section libraries used in lattice physics code. For few energy group homogenized macroscopic cross-section used in core analysis code, the covariance and sensitivity matrices are numerically evaluated, and then the cross-section table in core analysis are upgrade by cross-section adjustment technique. As a feasibility study, our proposed technique is verified for small core geometries based on the published data of BWR (Peach Bottom-2). Through the verification, it is confirmed that core characteristics (criticality and assembly-averaged relative power) during operation period can be improved by core analysis using adjusted few energy-group cross-sections.

　前のページ - 次のページ　

<遠藤 知弘>

MISC - 遠藤 知弘

<遠藤　知弘>

MISC - 遠藤　知弘