研究者詳細 - 遠藤　知弘

論文 - 遠藤　知弘

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Applicability of a reduced order model for a safety analysis code to statistical safety analysis 査読有り国際誌

Masaki Matsushita, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 57 巻 ( 12 ) 頁： 1307 - 1318 2020年12月

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 reduced order model (ROM) for a safety analysis code (RELAP/SCDAPSIM) is developed and is applied to the statistical safety analysis in order to reduce computational load. The ROM that reproduces output of the safety analysis code (e.g., the peak cladding temperature) is constructed using the singular value decomposition and the low-rank approximation for time-series data of a safety analysis code. A sampling method that incorporates dissimilarity among input data is tested to generate training data for the ROM. A loss of coolant accident with multiple failures of safety functions in a BWR is considered as an accident scenario. Statistical results such as the average, the standard deviation, and the 95% cumulative probability value of safety parameters obtained by the ROM and those obtained by the safety analysis code are compared. The reference results obtained by the direct random sampling method with the safety analysis code are within the 95% confidence interval of the results obtained by ROM. Thus the applicability and effectiveness of the ROM to the statistical safety analysis are confirmed through numerical analysis.

DOI： 10.1080/00223131.2020.1783382

Implementation of the unscented transformation with low rank approximation in uncertainty analysis during large-break loss of coolant accident 査読有り国際誌

Basma Foad, Akio Yamamoto, Tomohiro Endo

Annals of Nuclear Energy 146 巻頁： 107614 2020年10月

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.

The Low Rank Approximation (LRA) and Unscented Transform (UT) are integrated to produce a new algorithm having the capability to decrease the time required for the uncertainty quantification during Loss of coolant accident (LOCA) in Pressurized Water Reactors (PWR). The LRA is an efficient technique used in reducing computational cost due to its ability to perform dimensionality reduction by revealing the active or important degrees of freedom and calculate the basis of the so-called active subspace basing on the Singular Value Decomposition (SVD). For further reduction in the computational time; the UT algorithm is also implemented to generate a set of sigma points, these sigma points are the representatives of the whole probability distribution (the UT is restricted to Gaussian distribution). The main safety parameter is the maximum cladding temperatures during the accident which are computed by ATHLET thermal-hydraulic code. The reactivity coefficients and the covariance matrix are calculated using the SCALE 6.2 code. The present calculation model has 14-dimensions, therefore the number of sigma points needed for the SVD/UT technique is 29, and can be minimized to 5 sigma points only if the LRA/UT is used where two singular values are sufficient to reproduce/span the space thanks to the strong correlations between the reactivity coefficients.

DOI： 10.1016/j.anucene.2020.107614

Uncertainty and regression analysis of the MSLB accident in PWR based on unscented transformation and low rank approximation 査読有り国際誌

Basma Foad, Akio Yamamoto, Tomohiro Endo

Annals of Nuclear Energy 143 巻頁： 107493 2020年8月

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.

The present studies focus on the quantification of uncertainty during the main steam line break accident scenario (MSLB) in PWR, assuming that there is a failure on the feed-water regulating valve of the broken steam generator. The scenario is characterized by the associated positive Doppler and coolant density reactivities which bring the core back to critical (return-to-power). Accordingly, the input uncertainty parameters are the Doppler and coolant density reactivities taking into account the correlation matrix among the input parameters, which is calculated by SCALE 6.2 code. The main safety parameters are the maximum cladding surface temperatures and average core power during the accident which are computed by ATHLET thermal-hydraulic code. The sampling-based uncertainty technique is considered to be the most dependable technique which can be applicable to any code, however it is computationally expensive. Therefore, it is important to develop efficient techniques which are capable of reducing the calculation time. The first approach is the SVD-UT where the Unscented Transform (UT) algorithm and singular value decomposition (SVD) are combined to generate a minimal sample points. In addition, due to the strong correlation between the input reactivities, the computational time can be further reduced by implementing the Low Rank Approximation (LRA) and revealing the active subspace.

DOI： 10.1016/j.anucene.2020.107493

Efficient uncertainty quantification for PWR during LOCA using unscented transform with singular value decomposition 査読有り国際誌

Basma Foad, Akio Yamamoto, Tomohiro Endo

Annals of Nuclear Energy 141 巻頁： 107341 2020年6月

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 discusses one of the most important issues facing the regulatory body while performing the uncertainty analysis of the nuclear reactor parameter during accident conditions. This problem is the long computational time required by the statistical sampling methods to compute the uncertainty. We overcome this problem by introducing the Unscented Transform (UT) algorithm and singular value decomposition (SVD). Where both algorithms are combined (SVD/UT) to generate a set of sigma points, these sigma points are the representatives of whole probability distribution. The uncertainty quantification is performed during Loss of coolant accident in Pressurized Water Reactor (PWR), where the input variable of uncertainty is the coolant density reactivity. The SCALE 6.2 code is used for calculating the reactivity coefficients and the covariance matrix. The response variables are the peak cladding temperatures during the accident which are computed by ATHLET thermal-hydraulic code. The results obviously confirm the efficiency of the SVD/UT sampling in predicting the new mean values, and assure its ability to reduce the sampling size leading to a dramatic reduction of computational cost.

DOI： 10.1016/j.anucene.2020.107341

Impact of Various Parameters on Convergence Performance of CMFD Acceleration for MOC in Multigroup Heterogeneous Geometry 査読有り国際誌

Yoshiki Oshima, Tomohiro Endo, Akio Yamamoto, Yasuhiro Kodama, Yasunori Ohoka, Hiroaki Nagano

Nuclear Science and Engineering 194 巻 ( 6 ) 頁： 477 - 491 2020年6月

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.

The impact of various parameters in the coarse mesh finite difference (CMFD) acceleration method on overall convergence behavior is investigated through numerical calculations using the method of characteristics (MOC). Four parameters appearing in the CMFD acceleration with MOC, i.e., scalar flux distribution in flat flux regions (FFRFlux), the scalar flux distribution in CMFD meshes (CMFDFlux), homogenized cross sections (HXSs) in CMFD meshes, and current correction factors (CCFs), are considered. Parts of these four parameters are fixed to the converged values throughout iterations in order to estimate their impact on convergence. Numerical calculations are carried out for Korea Advanced Institute of Science and Technology’s (KAIST’s) benchmark problem KAIST-2A, which is a heterogeneous and multigroup problem, and the number of outer iterations to reach convergence is evaluated. The impact of geometric heterogeneity and cross-section homogenization in the CMFD acceleration has not been considered in linearized Fourier analysis so far. The calculation results indicate that (1) convergence of HXS has little impact on the overall convergence, (2) convergence of FFRFlux is dominant followed by CCF when a CMFD mesh is optically thin, and (3) convergence of FFRFlux is dominant when a CMFD mesh is optically thick and contains many flat flux regions.

DOI： 10.1080/00295639.2020.1722512

Application of the multigrid amplitude function method for time-dependent MOC based on the linear source approximation 査読有り国際誌

Kosuke Tsujita, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 57 巻 ( 6 ) 頁： 646 - 662 2020年6月

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 numerical scheme for time-dependent MOC calculations is proposed. In the present scheme, one of the most successful factorization method, the multigrid amplitude function (MAF) method, is employed to achieve faster computation with the minimum degradation for the temporal integration of the scalar flux. In addition, the MAF method is re-derived based on the linear source approximation, which is not applied for time-dependent MOC calculations in the past studies as far as the authors’ knowledge, to reduce the spatial discretization error with the coarser flux region separation. The accuracy and computational time of the present scheme are evaluated through the calculation of the TWIGL and the C5G7-TD 2D benchmark problems. The present calculation results show that the present scheme is 6.2 times faster than the conventional method while achieving the same accuracy in the C5G7-TD benchmark problem.

DOI： 10.1080/00223131.2019.1709993

Development of a reduced order model for severe accident analysis codes by singular value decomposition aiming probabilistic safety margin analysis 査読有り国際誌

Masaki Matsushita, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 57 巻 ( 5 ) 頁： 573 - 589 2020年5月

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.

Severe accident codes (e.g. MAAP, RELAP, and MELCORE) model various physical phenomena during severe accidents. Many analyses using these codes for safety margin evaluation are impractical due to large computational costs. Surrogate models have an advantage of quickly reproducing multiple results with a low computational cost. In this study, we apply the singular value decomposition to the time-series results of a severe accident code to develop a reduced order modeling (ROM). Using the ROM, the probabilistic safety margin analysis for the station blackout with a total loss of feedwater capabilities at a boiling water reactor is carried out. The dominant parameters to the accident progression are assumed to be the down-time and the recovery-time of the reactor core isolation cooling system, and decay heat. To reduce the number of RELAP5/SCDAPSIM analyses while maintaining the prediction accuracy of ROM, we develop a data sampling method based on adaptive sampling, which selects the new sampling data based on the dissimilarity with the existing training data for ROM. Our ROM can rapidly reproduce the time-series results and can estimate core conditions. By reproducing multiple results by ROM, a time-dependent core damage probability distribution is calculated instead of the direct use of RELAP5/SCDAPSIM.

DOI： 10.1080/00223131.2019.1699190

Subcriticality Measurement using Time-Domain Decomposition-Based Integral Method for Simultaneous Reactivity and Source Changes 査読有り国際誌

Tomohiro Endo, Asahi Nonaka, Sho Imai, Akio Yamamoto, Atsushi Sakon, Kengo Hashimoto

Journal of Nuclear Science and Technology 57 巻 ( 5 ) 頁： 607 - 616 2020年5月

担当区分：筆頭著者,　責任著者記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis, Ltd.

To estimate the subcriticality in dollar units for an arbitrary state-change, the time-domain decomposition-based integral method (TDDI) is proposed using the point kinetics theory based on the fundamental mode approximation. In a general transient subcritical system, reactivity, neutron source intensity, and point kinetics parameters can vary simultaneously. Furthermore, the state-change may not necessarily be a stepwise change. For such a transient, the TDDI method can estimate the subcriticality after the transient using only the time variation of the neutron count rate. Therefore, the proposed method is useful to approximately estimate the subcriticality in a system where a detailed core configuration is unknown. To investigate the applicability of the TDDI method, transient experiments with simultaneous reactivity and source changes or to two successive safety rods dropping were performed at the Kindai University Training and Research Reactor (UTR-KINKI). By comparing with reference values using excess reactivity and control rod worth, it was validated that the subcriticality values obtained by the TDDI method better agree with the reference values than the previous integral method.

DOI： 10.1080/00223131.2019.1706658

その他リンク： https://doi.org/10.1080/00223131.2019.1706658

Loading pattern optimization for a PWR using Multi-Swarm Moth Flame Optimization Method with Predator 査読有り国際誌

Satomi Ishiguro, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 57 巻 ( 5 ) 頁： 523 - 536 2020年5月

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.

This paper aims to propose a new methodology for optimizing fuel loading pattern in a nuclear reactor which is important for its higher safety and economic efficiency. Previous researches have proposed various methodologies to decide better loading patterns automatically. However, the processes still require manual operations of engineers to automatically design actual loading patterns. Swarm intelligence algorithm has currently gained interest as a solution to seek the patterns. Although these methodologies generate better patterns, they sometimes struggle with getting out from local optima and fails to complete the optimization. Large and multimodal solution space sometimes captures worse solutions due to local optima. The conventional methodologies struggle with setting proper parameters to get out from local optima. This research focuses on Multi-Swarm Moth Flame Optimization with Predator (MSMFO-P), an improved Moth Flame Optimization (MFO) by applying the concepts of predator and multi-swarm, as new methodologies. The method of MSMFO-P was applied to solve a loading pattern problem and compared with the conventional optimization methods such as simulated annealing (SA), Hybrid genetic algorithm (GA), and particle swarm optimization (PSO). The results of our experimental works indicated that MSMO-P generates better loading patterns than the conventional methodologies.

DOI： 10.1080/00223131.2019.1700844

Nuclear data-induced uncertainty quantification of prompt neutron decay constant based on perturbation theory for ADS experiments at KUCA 査読有り国際誌

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

Journal of Nuclear Science and Technology 57 巻 ( 2 ) 頁： 196 - 204 2020年2月

担当区分：筆頭著者,　責任著者記述言語：英語掲載種別：研究論文（学術雑誌）出版者・発行元：Taylor & Francis, Ltd.

In experimental benchmarks of the accelerator-driven system (ADS) conducted at the Kyoto University Critical Assembly (KUCA), the prompt neutron decay constant α was measured using two types of pulsed neutron sources, i.e. a D-T neutron source and a spallation neutron source driven by a 100-MeV proton beam. The measurement results of α are useful information to validate the numerical results predicted by the prompt ω-eigenvalue calculation. In this study, the numerical analysis of α using a multi-energy group SN neutron transport code was carried out for the uranium-lead zoned experimental cores. To reduce the discretization error owing to the deterministic code, the KUCA geometry was modelled in detail as a three-dimensional heterogeneous plate-by-plate geometry, and an improved variant of EON quadrature was utilized. In addition, the sensitivity coefficients of α with respect to nuclear data were efficiently evaluated by first-order perturbation theory, followed by nuclear data-induced uncertainty quantification based on the 56 neutron-energy group SCALE covariance library. Consequently, the numerical results of α were validated successfully by the experimental results of the pulsed neutron source method, compared with the range of the nuclear data-induced uncertainties.

DOI： 10.1080/00223131.2019.1647893

Numerical benchmark problem of solid-moderated enriched-uranium-loaded core at Kyoto university critical assembly 査読有り国際誌

Go Chiba, Tomohiro Endo

Journal of Nuclear Science and Technology 57 巻 ( 2 ) 頁： 187 - 195 2020年2月

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.

Useful and valuable measurement data obtained at the solid-moderated core for the development of accelerator-driven systems (ADSs) have been accumulated at the Kyoto University Critical Assembly (KUCA), and some of them have been open to the public. In order to efficiently utilize these data, experimental analyses with deterministic calculation procedures are helpful. In the present manuscript, a numerical benchmark problem is established. This benchmark problem can be utilized by users of the ADS-related measurement data obtained at the KUCA A-core to verify their own numerical tools devoted to experimental analyses. Material and geometrical specifications with reference solutions obtained by a continuous-energy Monte Carlo code MVP-II are provided.<br />
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In addition, numerical results obtained by a deterministic code system CBZ are also presented as an example. Through careful investigation about discretization on space and angle, guideline for proper discretization is provided. The CBZ results tend to underestimate the reference Monte Carlo solutions about 0.5%∆k/kk', and calculations of simplified core models suggest that this is caused by neutron leakage treatment in finite systems or resonance self-shielding treatment in CBZ.

DOI： 10.1080/00223131.2019.1647892

Real-time subcriticality monitoring system based on a highly sensitive optical fiber detector in an accelerator-driven system at the Kyoto University Critical Assembly 査読有り国際誌

Watanabe Kenichi, Yamanaka Masao, Endo Tomohiro, Pyeon Cheol Ho

Journal of Nuclear Science and Technology 57 巻 ( 2 ) 頁： 136 - 144 2020年2月

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 real-time subcriticality monitoring system for accelerator-driven system (ADS) is developed at the Kyoto University Critical Assembly (KUCA). The monitoring system consists of a highly sensitive optical fiber-based neutron detector, which combines a large number of small pieces of LiF/Eu:CaF2 eutectic scintillators with a wavelength-shifting fiber, and a digital signal processing system. The fabricated detector is sufficiently small for insertion into slight gaps between the fuel assemblies of the KUCA core and clearly reveals a peak shape indicating neutron signals in the pulse height spectrum. As subcriticality determination procedures, the system employs the Rossi-α, α-fitting, and area ratio methods. The subcriticality experimental results and their related parameters are compared with numerical results. Although the area ratio method slightly underestimates the parameters, overall, the estimated parameters agree well with the calculated ones. Finally, the developed system is found to evaluate the subcriticality within an accuracy of approximately 20% for the effective neutron multiplication factor keff ranges between 0.97 and 0.93, which are used in typical ADS experiments.

DOI： 10.1080/00223131.2019.1647895

Experimental analyses of βeff/Λ in accelerator-driven system at Kyoto University Critical Assembly 査読有り国際誌

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

Journal of Nuclear Science and Technology 57 巻 ( 2 ) 頁： 205 - 215 2020年2月

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.

The capability of βeff/Λ obtained by λ-mode and ω-mode calculations is examined for target subcriticality in the accelerator-driven system through a comparison with that obtained in the pulsed-neutron source (PNS) experiments at the Kyoto University Critical Assembly. Directly measured results of βeff/Λ, α and ρ_$ in the PNS experiments are validated by varying the experimental conditions: the external neutron source, detector position, detector type and subcriticality ranging between 500 and 7500 pcm. The numerical analyses of βeff/Λ are conducted by using MCNP6.1 together with ENDF/B-VII.1 for λ-mode calculations and PARTISN (SCALE6.2.2 with ENDF-VII.1 for the effective cross sections) for both the λ-mode and the ω-mode calculations. The comparison between calculated and measured βeff/Λ with varying subcriticality shows good agreement between the experiments and the ω-mode calculations, although a difference is observed between the experiments and the λ-mode calculations.

DOI： 10.1080/00223131.2019.1666057

A New Interpretation of Discontinuity Factor 査読有り国際誌

Akio Yamamoto, Tomohiro Endo

Nuclear Science and Engineering 193 巻 ( 9 ) 頁： 991 - 997 2019年9月

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 interpretation of the discontinuity factor (DF) for scalar flux, partial current, and angular flux is discussed. Conventionally, the DF is considered as the discontinuous condition of scalar flux, partial current, or angular flux at an interface. In the new interpretation, the DF is considered as the refractive index of materials for partial current or angular flux that conserves odd-parity or odd-moment angular flux at an interface of different materials. It is related to the transmission and reflection of partial current or angular flux at an interface where different materials are adjacent. Using the present interpretation, a fundamental issue of neutron balance (i.e., artificial loss or production of neutrons at an interface due to discontinuous condition), which would appear in the conventional interpretation of DF, can be resolved.

DOI： 10.1080/00295639.2019.1579514

First nuclear transmutation of Np237 and Am241 by accelerator-driven system at Kyoto University Critical Assembly 査読有り国際誌

Cheol Ho Pyeon, Masao Yamanaka, Akito Oizumi, Masahiro Fukushima, Go Chiba, Kenichi Watanabe, Tomohiro Endo, Wilfred G. Van Rooijen, Kengo Hashimoto, Atsushi Sakon, Naoto Aizawa, Yasutoshi Kuriyama, Tomonori Uesugi, Yoshihiro Ishi

Journal of Nuclear Science and Technology 56 巻 ( 8 ) 頁： 684 - 689 2019年8月

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.

This study demonstrates, for the first time, the principle of nuclear transmutation of minor actinide (MA) by the accelerator-driven system (ADS) through the injection of high-energy neutrons into the subcritical core at the Kyoto University Critical Assembly. The main objective of the experiments is to confirm fission reactions of neptunium-237 (Np237) and americium-241 (Am241), and capture reactions of Np237. Subcritical irradiation of Np237 and Am241 foils is conducted in a hard spectrum core with the use of the back-to-back fission chamber that obtains simultaneously two signals from specially installed test (Np237 or Am241) and reference (uranium-235) foils. The first nuclear transmutation of Np237 and Am241 by ADS soundly implemented by combining the subcritical core and the 100 MeV proton accelerator, and the use of a lead-bismuth target, is conclusively demonstrated through the experimental results of fission and capture reaction events.

DOI： 10.1080/00223131.2019.1618406

Transport consistent diffusion coefficient for CMFD acceleration and comparison of convergence properties 査読有り国際誌

Akio Yamamoto, Tomohiro Endo, Akinori Giho

Journal of Nuclear Science and Technology 56 巻 ( 8 ) 頁： 716 - 723 2019年8月

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 diffusion coefficient for the coarse mesh finite difference (CMFD) acceleration is derived from the semi-analytic solution of one-group, one-dimensional, even-parity transport equation. The derived diffusion coefficient, i.e., the transport consistent diffusion coefficient (TCD), depends on the optical length of a mesh and shows similar behavior with the artificial grid diffusion (AGD) and the effective diffusion (EffD) coefficients for an optically thick mesh. Convergence properties of typical diffusion coefficients are evaluated using the linearized Fourier analysis. Analyses of the C5G7 3D benchmark problems with and without voided region are carried out to compare the convergence properties. The number of transport sweeps to reach convergence using TCD is smaller than that using EffD or AGD.

DOI： 10.1080/00223131.2019.1618405

A simple treatment of increased gap due to fuel assembly bowing through correction of cross sections 査読有り国際誌

Akio Yamamoto, Tomohiro Endo, Hiroaki Nagano, Yasunori Ohoka, Kento Yamamoto

Journal of Nuclear Science and Technology 56 巻 ( 6 ) 頁： 471 - 478 2019年6月

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.

Increased fuel assembly gap due to bowing in commercial light-water reactors (LWRs) has an impact on local pin-power distribution due to increased local moderation. In order to consider the effect of increased assembly gap without explicit consideration of increased gap width, a correction method of cross sections in the gap region is proposed. In this cross section correction method, the average chord length of gap region is preserved to capture the effect of increased gap width. The validity of the present method is confirmed by verification calculations in a single assembly and 5 × 5 assembly geometries using the GENESIS code, which is a transport code based on the method of characteristics.

DOI： 10.1080/00223131.2019.1598509

Utilization of Regionwise Even-Parity Discontinuity Factor to Reduce Discretization Error of MOC 査読有り国際誌

Yamamoto Akio, Giho Akinori, Endo Tomohiro

Nuclear Science and Engineering 193 巻 ( 3 ) 頁： 253 - 268 2019年3月

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.

To reduce angular and spatial discretization error of the method of characteristics with a coarse calculation condition, the regionwise even-parity discontinuity factor (EPDF) for transport calculations is evaluated through an iterative procedure using only the regionwise scalar flux, i.e., without the odd-parity angular flux, the partial current, or the net current at the region boundary. The regionwise EPDF is evaluated in a single-assembly geometry with the reflective boundary condition. The evaluated EPDF is applied to a 2 × 2 colorset assembly and core configurations, and the performance is compared to that of the conventional superhomogenization (SPH) method. The calculation results indicate that (1) no convergence issue is observed during the iteration process to estimate the EPDF, (2) the performance of the regionwise EPDF is better than that of the conventional SPH method, and (3) the normalization of the EPDF is necessary to incorporate different surface scalar flux levels among different types of fuel assemblies.

DOI： 10.1080/00295639.2018.1516961

Inverse estimation methods of unknown radioactive source for fuel debris search 招待有り査読有り国際誌

Shinji Sugaya, Tomohiro Endo, Akio Yamamoto

Annals of Nuclear Energy 124 巻頁： 49 - 57 2019年2月

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.

To identify the distribution of fuel debris remaining in the reactor vessel and/or the containment vessel of Fukushima Daiichi NPS, we focused on the inverse estimation of radioactive source distribution using the measured radiation counts. The Maximum Likelihood-Expectation Maximization (ML-EM) and the Moore-Penrose Matrix Inverse (MPMI) methods are examined. The ML-EM method has been used for the image reconstruction of computed tomography, and the MPMI method is one of the solution methods for simultaneous linear equations with the underdetermined condition. A simple calculation model simulating a containment vessel was constructed including detectors and radiation sources. In an actual situation, a sufficient number of radiation measurement positions would not be available owing to the complexity of structures inside the containment vessel. Thus, the number of radiation measurement points (number of constraints) is smaller than that of radiation source positions. It means that an underdetermined inverse problem should be solved. The detection probability of radiation (neutron or photon) is calculated by the adjoint transport calculation since the detection probability is used as the coupling coefficient between radiation counts at a detector and a radioactive source. The result of estimation using the ML-EM or the MPMI method indicates that the accuracy of estimation depends on the distance between a radiation source and a detector, and measurement positions of radiation count. The ML-EM and the MPMI methods show different prediction accuracy depending on the prediction condition. It is found that reasonable prediction accuracy would be obtained when the detectors are placed at the vicinity of radiation sources of interest.

DOI： 10.1016/j.anucene.2018.09.022

Flux Region Assignment Method Using Ray Trace Information for the Method of Characteristics to Improve Cache Efficiency 査読有り国際誌

Akio Yamamoto, Akinori Giho, Tomohiro Endo

Nuclear Science and Engineering 192 巻 ( 3 ) 頁： 240 - 253 2018年12月

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 flux region assignment algorithm to increase cache efficiency for the method of characteristics (MOC) is proposed. In order to minimize the stride of memory access, flux region identifications are assigned based on the ray trace sequence during the MOC calculation. The present method is implemented in the three-dimensional transport code GENESIS and its performance is confirmed through verification calculations ranging from single pressurized water reactor (PWR) fuel assembly to PWR full-core benchmark problems. Quantitative comparison of cache efficiency is carried out and the present method shows improved cache efficiency, which results in a reduction in computation time. The present method can reduce computational time by improving cache efficiency while suppressing memory requirement.

DOI： 10.1080/00295639.2018.1501978

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