MISC - ENDO Tomohiro

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

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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

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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

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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

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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

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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.

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

Trans. Am. Nucl. Soc.   Vol. 115   page： 1085 - 1087   2016.11

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

The random sampling method is applied to neutron shielding calculation using ANISN/TRANSX in order to estimate uncertainties of neutron flux and activation in a structure material (iron). The calculation results indicate that the magnitude of uncertainty of activation of Co59 due to cross section is approximately 10% inside structural material, which is not very large when attenuation of neutron flux is taken into account.

Prediction on Underestimation of Statistical Uncertainty in Monte Carlo Eigenvalue Calculation for Two-dimensional Heterogeneous Color Set Geometry Reviewed International journal

K. Hayashi, T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc.   Vol. 115   page： 1213 - 1216   2016.11

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

Theoretical formula for behavior of noise term is derived, which is necessary for prediction of underestimation ratio of statistical uncertainty in Monte Carlo eigenvalue calculation. By using the proposal theoretical formula, underestimation ratio of flux of each fuel cell in 2×2 UO2-MOX assembly geometry is predicted and compared with the reference results obtained by 1000 GMVP runs. The prediction results show reasonable agreement with the reference regarding to relative spatial shape, which suggest the validity of the theoretical formula of the underestimation ratio and the noise term. However, there is considerable difference on the absolute values of the underestimation ratio and investigation of the causes of the difference would be a future task.

Analysis and Interpretation of the KUCA ADS Benchmarks with Deterministic Analysis Codes Reviewed International journal

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

Proc. PHYSOR2016     page： 4200 - 4216   2016.5

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

This manuscript discusses the analysis of the KUCA-A ADS with deterministic codes (specifically diffusion theory). The KUCA-A ADS is a small, thermal-spectrum ADS driven by 100MeV protons. Several sets of experiments are available as an international benchmark. In our analysis, we calculated the multiplication factor, prompt neutron decay constant (α-eigenvalue), performed the analysis of an irradiation experiment, and used time-dependent diffusion theory to analyze pulsed neutron operation of the KUCA-A ADS in 3D. 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.

Variance Reduction Factor Calculations for Neutronics Parameters of Accelerator-Driven System Reviewed International journal

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

Proc. PHYSOR2016     page： 1661 - 1668   2016.5

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

In order to identify nuclear data on which further improvement is effective to reduce nuclear data-induced uncertainties of accelerator-driven system (ADS) neutronics parameters, variance reduction factors (VRFs) for keff, beff and coolant void reactivity are calculated. Obtained VRFs suggest that improvement of plutonium-238 nuclear data is quite effective to reduce the uncertainties of keff and beff during the operation of ADS, and that improvement of lead-206 elastic and inelastic scattering cross sections is also important to reduce the uncertainties of coolant void reactivity. Important energy ranges of these nuclear data are also specified by VRFs.

GENESIS - A Transport Solver in Three-Dimensional Heterogeneous Geometry Based on the LEAF Method Reviewed International journal

A. Yamamoto, A. Giho, T. Endo

Proc. PHYSOR2016     page： 4080 - 4106   2016.5

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

A heterogeneous transport solver in three dimensional geometry, GENESIS, is being developed incorporating recent developments of the method of characteristics (MOC) in three dimensional geometry. The Legendre Expansion of Angular Flux (LEAF) method is implemented in the GENESIS code, in which neutron transport is calculated in two-dimensional "characteristics planes" rather than in one-dimensional characteristics lines adopted in the conventional approach of 3D MOC. Unlike the planner MOC method that combines 2D 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 planner MOC method. The GENESIS code can handle flexible shape of objects in rectangular or hexagonal geometry. Two-level, multi-group generalized coarse mesh rebalance (GCMR) 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.

Theoretical Expression of Area Ratio Method using Detected-Neutron Multiplication Factor Reviewed International journal

T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc.   Vol. 113   page： 1208 - 1211   2015.11

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

In the present paper, the area ratio method is expressed by the detected-neutron multiplication factor kdet. In addition, an appropriate detector position is discussed by investigating the difference between kdet and the effective neutron multiplication factor keff.

Application of Data Assimilation based on Bayesian Theory to Subcriticality Measurements using Area Ratio Method Reviewed International journal

K. Maeno, T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc.   Vol. 113   page： 1282 - 1286   2015.11

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

Underestimation of Statistical Uncertainty of Monte Carlo Method with Non-Analog of Fission Source Sampling Reviewed International journal

K. Hayashi, T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc.   Vol. 113   page： 1153 - 1157   2015.11

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

Uncertainty Estimation of Analysis Model using the Data Assimilation Method Reviewed International journal

K. Kinoshita, T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc.   Vol. 113   page： 1141 - 1143   2015.11

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Development of a Simplified Estimation Method on Severe Accident Progression in PWR for Education Reviewed International journal

S. Otsuki, T. Endo, A. Yamamoto

Trans. Am. Nucl. Soc.   Vol. 113   page： 855 - 858   2015.11

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Application of Partially-Converged Solution of Assembly Calculation for Core Sensitivity Analysis based on Reduced Order Modeling Reviewed International journal

R. Katano, A. Yamamoto, T. Endo

Trans. Am. Nucl. Soc.   Vol. 113   page： 1161 - 1164   2015.11

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Discontinuity Factors for Simplified P3 Theory Reviewed

A. Yamamoto, T. Endo

Proc. of Reactor Physics Asia 2015   Vol. CD-ROM   2015.9

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Reactor Physics Activities in Nagoya University Reviewed

A. Yamamoto, T. Endo

Proc. of Reactor Physics Asia 2015   Vol. CD-ROM   2015.9

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Development of New Statistical Geometry Model using the Delta-tracking Method Reviewed

T. Koide, T. Endo, A. Yamamoto

Proc. of Reactor Physics Asia 2015   Vol. CD-ROM   2015.9

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Development of Core Sensitivity Analysis Based on Reduced-Order Modeling using Assembly Calculations Reviewed International journal

R. Katano, A. Yamamoto, T. Endo

Trans. Am. Nucl. Soc.   Vol. 112   page： 715 - 718   2015.6

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