Faculty Profiles - ENDO Tomohiro

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Explicit Estimation of Higher Order Modes in Fission Source Distribution of Monte-Carlo Calculation Reviewed International journal

A. Yamamoto, K. Sakata, T. Endo

Proc. M&C2013 page： 2814 - 2827 2013.5

Language：English Publishing type：Research paper, summary (international conference)

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 Reviewed International journal

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

Proc. M&C2013 page： 1826 - 1838 2013.5

Language：English Publishing type：Research paper, summary (international conference)

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 Reviewed International journal

K. Tsujita, T. Endo, A. Yamamoto

Proc. M&C2013 page： 1839 - 1852 2013.5

Language：English Publishing type：Research paper, summary (international conference)

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 Reviewed International journal

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

Proc. Int. Congress on Advances in Nuclear Power Plants (ICAPP2013) page： 877 - 883 2013.4

Authorship：Corresponding author Language：English Publishing type：Research paper, summary (international conference)

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.

Correction Technique for Coarse Group Cross Sections Considering Spectral Interference Effect on Pin-by-Pin BWR Core Analysis Reviewed International journal

T. Fujita, T. Endo, A. Yamamoto

Proc. Int. Congress on Advances in Nuclear Power Plants (ICAPP2013) page： 936 - 945 2013.4

Correction of Spectral Interference Effect on Pin-by-Pin BWR Core Analysis Reviewed International journal

T. Fujita, T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc. Vol. 107 page： 1141 - 1143 2012.11

Subcriticality Measurement by Neutron Source Multiplication Method with Detected-Neutron Multiplication Factor Reviewed International journal

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

Trans. Am. Nucl. Soc. Vol. 107 page： 648 - 651 2012.11

Efficient Calculation Scheme with Preservation of Transmission Probabilities in the Method of Characteristics Reviewed International journal

M. Tabuchi, N. Sugimura, A. Yamamoto, T. Endo

Trans. Am. Nucl. Soc. Vol. 107 page： 1105 - 1107 2012.11

Language：English Publishing type：Research paper, summary (international conference) Publisher：American Nuclear Society

In this study, an efficient calculation scheme for MOC is investigated. The equation for the effective length of a path line is derived based on an assumption that the lengths are linearly distributed in a strip. By using the effective length, the transmission probabilities in each strip can be preserved. Through the verification calculations, the accuracy of the present method can be confirmed, and efficiency of the MOC calculation is expected to be improved by the present method.

Higher Order Treatment on Temporal Derivative of Angular Flux for Time-Dependent MOC Reviewed International journal

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

Trans. Am. Nucl. Soc. Vol. 107 page： 1101 - 1104 2012.11

Kinetic Calculation Method in Space-Time Frame Using Characteristic Line Reviewed International journal

K. Tsujita, T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc. Vol. 106 page： 743 - 746 2012.6

Analysis of Erbia-loaded Critical Experiments in KUCA using AEGIS Cross Section Library Reviewed International journal

A. Yamamoto, T. Endo, X. Wu

Trans. Am. Nucl. Soc. Vol. 106 page： 715 - 718 2012.6

Estimation of Average Burnup of Damaged Fuels Loaded in Fukushima Dai-ichi Reactors by using the Cs134/Cs137 Ratio Method Reviewed International journal

T. Endo, S. Sato, A. Yamamoto

Proc. Physor2012 - Advanced in Reactor Physics - Linking research, Industry and Education page： 3246 - 3253 2012.4

Authorship：Lead author,　Corresponding author Language：English Publishing type：Research paper, summary (international conference)

Cs134/Cs137 ratio method for measured radioactivities of 134Cs and 137Cs in contaminated soils within the range of 100 km from the Fukushima Dai-ichi nuclear power plants. As a result, the measured Cs134/Cs137 ratio from the contaminated soil is 0.996±0.07 as of March 11th, 2011. Based on the Cs134/Cs137 ratio method, the estimated burnup of damaged fuels is approximately 17.2±1.5 [GWd/tHM]. It is noted that the numerical results of various calculation codes (SRAC2006/PIJ, SCALE6.0/TRITON, and MVP-BURN) are almost the same evaluation values of Cs134/Cs137 ratio with same evaluated nuclear data library (ENDF-B/VII.0). The void fraction effect in depletion calculation has a major impact on 134Cs/137Cs ratio compared with the differences between JENDL-4.0 and ENDF-B/VII.0.

Uncertainty Estimation of Core Safety Parameters using Cross-Correlations of Covariance Matrix Reviewed International journal

A. Yamamoto, Y. Yasue, T. Endo, Y. Kodama, Y. Ohoka, M. Tatsumi

Proc. Physor2012 - Advanced in Reactor Physics - Linking research, Industry and Education page： 2880 - 2895 2012.4

Language：English Publishing type：Research paper, summary (international conference)

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

Development of a Lattice Physics Code for Sensitivity Analysis Based on Generalized Perturbation Theory Reviewed International journal

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

Trans. Am. Nucl. Soc. Vol. 105 page： 851 - 854 2011.10

Evaluation of Correlations among Uncertainties of Core Neutronic Parameters in Infinite Slab Geometry Reviewed International journal

Y. Yasue, T. Endo, A. Yamamoto, Y. Kodama, Y. Ohoka, M. Tatsumi

Trans. Am. Nucl. Soc. Vol. 105 page： 486 - 488 2011.10

Assembly Discontinuity Factor for Angular Flux in Transport Calculation Reviewed International journal

A. Yamamoto, T. Endo

Trans. Am. Nucl. Soc. Vol. 105 page： 862 - 864 2011.10

A Unified Numerical Algorithm for Space-Dependent Kinetic Equation Invited Reviewed International journal

T. Endo, Y. Ban, A. Yamamoto

Trans. Am. Nucl. Soc. Vol. 104 page： 865 - 867 2011.6

A Derivation of Discontinuity Factor for Angular Flux in Integro-Differential Transport Equation Reviewed International journal

A. Yamamoto, T. Endo, Y. A. Chao

Trans. Am. Nucl. Soc. Vol. 104 page： 815 - 817 2011.6

Search Strategy of Detector Position for Neutron Source Multiplication Method by using Detected-Neutron Multiplication Factor Invited International journal

T. Endo

Transactions of Korean Nuclear Society 2011 Spring Meeting page： 131 - 132 2011.5

Authorship：Lead author,　Corresponding author Language：English Publishing type：Research paper, summary (international conference) Publisher：Korean Nuclear Society

In this paper, based on the idea of kdet, it is clear that correlation factors of the NSM method consist of the sourc-flux correction factor fs to fix the difference of non-fission component of neutron count rate, and the conversion factor fc to convert from kdet to the effective multiplication factor keff. Through the numerical analysis of three-dimensional two-group transport calculations for simple geometry, appropriate detector positions where kdet=keff could be find out for only one certain subcritical state, e.g., the reference state. By putting a neutron detector at such a particular position, the target neutron multiplication factors can be well estimated even without any corrections.

Effect of Uncertainty of Planned Cycle Length in Multicycle Fuel Optimization Reviewed International journal

K. Ohori, T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc. Vol. 103 page： 707 - 710 2010.11

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