Faculty Profiles - ENDO Tomohiro

MISC - ENDO Tomohiro

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Application of the GENESIS Code to the Kobayashi 3D Benchmark Problem Reviewed International journal

A. Yamamoto, A. Giho, T. Endo

Trans. Am. Nucl. Soc. Vol. 117 page： 1403 - 1406 2017.10

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

The GENESIS code, which is a transport code based on the LEAF method for a three-dimensional heterogeneous geometry, is applied to the Kobayashi 3D benchmark problem that is characterized by deep penetration and streaming through a void region.

Estimation of Modeling Approximation Error of Core Analysis using the Surrogate Model with Kriging Reviewed International journal

T. Hanai, T. Endo, Y. Kodama, Y. Ohoka, A. Yamamoto

Trans. Am. Nucl. Soc. Vol. 117 page： 1269 - 1272 2017.10

Application of the Bias Factor Method using the Random Sampling Technique for Prediction Accuracy Improvement of Neutronics Parameters of BWR Reviewed International journal

M. Ito, T. Endo, A. Yamamoto, Y. Kuroda, T. Yoshii

Trans. Am. Nucl. Soc. Vol. 117 page： 804 - 807 2017.10

On prediction accuracy of neutronics parameters of accelerator-driven system Reviewed International journal

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

EPJ Web of Conferences Vol. 146 page： 09005 2017.9

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

Nuclear data-induced uncertainty of neutronics parameters of one specific ADS design is quantified. The nuclear data adjustment method with available integral data is employed to reduce the uncertainties, and usefulness of these integral data is investigated. Numerical results reveal that the uncertainty reduction by the present nuclear data adjustment is insignificant and restrictive. Future perspecitives are also provided.

DOI： 10.1051/epjconf/201714609005

Scopus

Estimation of Reactivity, Neutron Generation Time, and Effective Delayed Neutron Fraction using Extended Kalman Filter Reviewed International journal

T. Ikeda, T. Endo, A. Yamamoto

Proc. Reactor Physics Asia 2017 Vol. CD-ROM page： CD-ROM 2017.8

Uncertainty Quantification due to Modeling Approximation Error of Core Analysis using the Kriging Method Reviewed International journal

T. Hanai, T. Endo, Y. Kodama, Y. Ohoka, A. Yamamoto

Proc. Reactor Physics Asia 2017 Vol. CD-ROM 2017.8

Application of Bias Factor Method using Random Sampling Technique for Prediction Accuracy Improvement of Critical Eigenvalue of BWR Reviewed International journal

M. Ito, T. Endo, A. Yamamoto, Y. Kuroda, T. Yoshii

Proc. ICAPP2017 page： 1709 - 1716 2017.4

Language：English Publishing type：Research paper, summary (international conference) Publisher：Atomic Energy Society of Japan

The bias factor method based on the random sampling technique is applied to the benchmark problem of Peach Bottom Unit 2. Validity and availability of the present method, i.e. correction of calculation results and reduction of uncertainty, are confirmed in addition to features and performance of the present method. In the present study, core characteristics in cycle 3 are corrected with the proposed method using predicted and "measured" critical eigenvalues in cycles 1 and 2. As the source of uncertainty, variance-covariance of cross sections is considered. The calculation results indicate that bias between predicted and measured results, and uncertainty owing to cross section can be reduced. Extension to other uncertainties such as thermal hydraulics properties will be a future task.

Uncertainty Quantification of Activation Due to Cross Section Data in Neutron Shielding Calculation Reviewed International journal

K. Yokoi, T. Endo, A. Yamamoto, R. Mizuno, Y. Kimura

Proc. ICAPP2017 page： 2152 - 2156 2017.4

Language：English Publishing type：Research paper, summary (international conference) Publisher：Atomic Energy Society of Japan

Quantification of activation in structure materials around a nuclear reactor using results of neutron shielding calculations is necessary for efficient planning of decommissioning. In this study, the random sampling method is applied to the neutron shielding calculation in order to estimate uncertainties of neutron flux and activation of Co59 in structure materials due to cross section covariance. The calculation results indicate that the magnitude of uncertainty of activation of Co59 due to cross section data is approximately 15~30% inside structure materials.

Evaluation of the n/γ Discrimination Performance of the Neutron Detector with Eu Doped TRUST-LiCaAlF6 Reviewed International journal

M. Maeno, T. Endo, A. Yamamoto

Proc. ICAPP2017 page： 1510 - 1516 2017.4

Authorship：Corresponding author Language：English Publishing type：Research paper, summary (international conference) Publisher：Atomic Energy Society of Japan

The n/γ discrimination performance of the prototype neutron detector using Eu doped TRUST-LiCaAlF6 is quantitatively investigated, in order to confirm its applicability to the subcriticality monitoring in the decommissioning works at Fukushima Daiichi Nuclear Power Plant units 1-3, especially focusing on removal of fuel debris. In this paper, the experimental results of pulse-height distribution measurement using Cf252 at the Nagoya University Cobalt 60 irradiation facility are shown. Additionally, in order to investigate the applicability to the reactor noise analysis, preliminary experimental results of the Feynman-α method are also shown.

Effective Use of Engineering Reactor Simulator for Education of Nuclear Safety Reviewed International journal

A. Yamamoto, T. Endo

Proc. ICAPP2017 page： 2065 - 2069 2017.4

Language：English Publishing type：Research paper, summary (international conference) Publisher：Atomic Energy Society of Japan

Effective use of a reactor simulator to lecture nuclear safety for undergraduate/master's course students majoring nuclear engineering is discussed. Various scenarios from normal operating to severe accident conditions are used for an exercise of a few days' course. Through discussions on plant behaviors of various scenarios within a small group and among class members, participants can understand physics behind responses of a nuclear power plant.

Development of Dynamic Probabilistic Risk Assessment Model for PWR using Simplified Plant Simulation Method Reviewed International journal

S. Otsuki, T. Endo, A. Yamamoto

Proc. ICAPP2017 page： 1737 - 1742 2017.4

Language：English Publishing type：Research paper, summary (international conference) Publisher：Atomic Energy Society of Japan

The CMMC (Continuous Markov Monte Carlo) coupling method is proposed for quantifying accident scenarios that have time-dependence in accident progression and inter-dependence of individual events. In order to analyze large number of samples within realistic computing time, we developed a simplified accident progression analysis model and coupled the simplified model to the CMMC method. Large number of samples for station black out in PWR were analyzed using the random sampling method based on the CMMC method considering uncertainty of decay heat from a core. As a result, uncertainties of event initiation time and FPs release ratio to the atmosphere are estimated.

Implementation of Fuel Burnup Sensitivity Calculation Capability into a Deterministic Reactor Physics Code System CBZ for Accelerator-Driven System Multi-Cycle Burnup Calculations Reviewed International journal

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

Proc. M&C 2017 page： P233 2017.4

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

A new capability of calculating fuel burnup sensitivity with the generalized perturbation theory is implemented into a deterministic reactor physics code system CBZ, which is under development at Hokkaido University. This capability is well verified through comparisons with reference sensitivities obtained by numerical differentiation. Nuclear data-induced uncertainties of two neutronics parameters, keff and coolant void reactivity, of an accelerator-driven system designed by the Japan Atomic Energy Agency are quantified using sensitivities calculated with the new version of CBZ, and effect of burnup term in the sensitivities is also quantified. On both keff and coolant void reactivity, cancellation between static and burnup components in nuclear data-induced uncertainties is observed; uncertainties become small if the burnup component is taken into account.

Inverse Estimation of Unknown Radioactive Source using Detection Probability by Adjoint Calculation Reviewed International journal

S. Sugaya, T. Endo, A. Yamamoto

Proc. M&C 2017 page： P411 2017.4

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

In order to investigate distribution of fuel debris remaining in the reactor containment vessel of Fukushima Daiichi NPS, we focused on the estimations based on the measured value of neutron counts. Maximum Likelihood-Expectation Maximization (ML-EM) method and Moore-Penrose Matrix Inverse (MPMI) method were examined. The ML-EM method is used for image reconstruction of Computed Tomography, and the MPMI method is one of the solution of the simultaneous linear equations. Concrete shielding were set between detectors and neutron sources in the calculation system, since where detector position in actual applications would be outside of the shielding such as the pedestal. Since sufficient number of count measurement positions would not be secured owing to the complexity of structures inside containment vessel, the number of measurement points were set to be smaller than that of neutron sources. Since the detection probability is used as the coupling coefficient between the radiation count and the radioactivity, the adjoint transport calculation is performed to obtain the detection probability. The detection probability calculated by adjoint neutron flux reproduced the physical trend expected from detector and neutron source positions. Result of estimation using this detection probability indicates that two methods gave reasonable neutron emission rate distribution that is close to the true value.

Benchmarks of Subcriticality in Accelerator-Driven System at Kyoto University Critical Assembly Reviewed International journal

C.H. Pyeon, M. Yamanaka, S.H. Kim, T.M. Vu, T. Endo, W.F.G. van Rooijen, G. Chiba

Proc. M&C 2017 page： P211 2017.4

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

Basic research on the accelerator-driven system is conducted with the combined use of U235-loaded core (the Kyoto University Critical Assembly) and the accelerators (100 MeV protons and 14 MeV neutrons). The results of experimental subcriticality are presented with a wide range of subcriticality level between near critical and 10000 pcm, as obtained by the measurement methods.

Uncertainty Quantification of Bell Factor for Sjöstrand Method using Random Sampling Method Reviewed International journal

T. Kimura, T. Endo, A. Yamamoto

Proc. M&C 2017 page： P206 2017.4

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

Sjöstrand Method is one of the subcriticality measurement techniques for the accelerator driven system (ADS). In this study, we investigated the uncertainty quantification of a spatial correction factor, Bell factor, due to cross-section data, using the random sampling method. As a result, the uncertainty of the Bell factor due to cross-section uncertainty is smaller than that of the subcriticality in dollar units, because the sensitivities of the Bell factor to cross-sections are cancelled between the area ratio and subcriticality.

Theoretical Discussion of Statistical Error for Variance-to-Mean Ratio Reviewed International journal

T. Endo, A. Yamamoto

Proc. M&C 2017 page： P305 2017.4

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

In the present paper, statistical error of variance-to mean ratio, or Y value in the Feynman-α method, is theoretically investigated to discuss the relationship among the statistical error of Y value, external neutron source strength, and measurement time. Practical theoretical formulae are derived to estimate the statistical error of Y value from a single measurement of reactor noise. The derived formulae clarify that the statistical error of Y value can be reduced by the total number of counting gate, or total measurement time, rather than the strength of external neutron source. Through an actual reactor noise experiment at the Kyoto University Criticality Assembly, the derived estimation formulae are validated.

Nuclear Data-Induced-Uncertainty Quantification of Neutron Multiplication Factor and Prompt Neutron Decay Constant for Pb-Bi loaded ADS Benchmark Problems at KUCA Reviewed International journal

T. Endo, A. Yamamoto

Proc. M&C 2017 page： P210 2017.4

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

By the sensitivity analysis using the SCALE6.2.1/TSUNAMI-3D and the random sampling method using the SCALE6.2.1/Sampler/NEWT/PARTISN, nuclear data-induced uncertainties of neutron multiplication factor keff and prompt neutron decay constant α are evaluated for Pb-Bi loaded Accelerator Driven System benchmark problems at Kyoto University Critical Assembly. As a result, the nuclear data-induced correlation between α and keff is strongly negative. It is supposed that the nuclear data-induced uncertainty of subcriticality is major contribution to the uncertainty of prompt neutron decay constant.

Development of GENESIS, a Three-dimensional Heterogeneous Transport Code based on the LEAF Method Reviewed International journal

A. Yamamoto, A. Giho, T. Endo

Proc. M&C 2017 page： P057 2017.4

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

The present 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, concept of the simplified Pn method is introduced in order to reduce storage of flux moments. Accuracy of the present method is verified through a benchmark problem. Next, iteration stability of the GENESIS code for highly voided condition, which would appear in the severe accident conditions, is discussed. Efficiency of the CMFD and GCMR acceleration methods are verified with various stabilization techniques. Use of the effective diffusion coefficient and the artificial grid diffusion (AGD) coefficients are effective to stabilize the acceleration calculation in highly voided conditions.

Comparison of Fuel Loading Pattern Optimization Results using Exhaustive Search for Fresh Fuels and Local Search for Burned Fuels Reviewed International journal

S. Ishiguro, T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc. Vol. 115 page： 1265 - 1267 2016.11

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

Optimization performances by combination of the exhaustive search for fresh fuel location and the MSLS are investigated. In the present study, the exhaustive binary swap with the exhaustive search for gadolinia fresh fuel location provides better result. Also, it is suggested that the current subspace divided by gadolinia fresh fuel location still shows multi-modality, thus the present local search method cannot find out an optimum solution in many cases. As a future task, more sophisticated method for subspace division, e.g., utilization of multi-stage divisions, would be necessary.

Uncertainty Quantification of Spatial Correction Factor for Sjöstrand Method due to Cross-section Data Reviewed International journal

T. Kimura, T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc. Vol. 115 page： 1081 - 1084 2016.11

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

We investigated the uncertainty quantification of the Bell factor, which is the spatial correction factor in subcriticality measurement using the Sjöstrand method, due to the uncertainty in cross-section data. As one of the feasibility studies, the uncertainty of the Bell factor is evaluated using the diffusion calculations for the simple one-dimensional slab geometry. As a result, the uncertainty of the Bell factor due to cross section is smaller than that of the subcriticality measurement in dollar units. In addition, as the difference in the calculation results between the area ratio and subcriticality in dollar units is smaller, the uncertainty of the Bell factor tends to become smaller. In future tasks, we will further investigate the same kind of uncertainty quantification for a practical ADS design with MA fuel, and the uncertainty quantification due to other factors, such as the delayed neutron fraction and the calculation method.

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