研究者詳細 - 遠藤　知弘

論文 - 遠藤　知弘

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Applicability of angular flux discontinuity factor preserving region-wise leakage for integro-differential transport equation 査読有り国際誌

Tatsuya Sakamoto, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 51 巻 ( 10 ) 頁： 1264 - 1273 2014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：Taylor & Francis, Ltd.

In the current core analysis, spatial homogenization is utilized to reduce the computational time. The discontinuity factor (DF) is one of the effective correction factors to reduce spatial homogenization error. The DF in diffusion equation is widely used on the other hand the DF in transport equation has not been put to practical use although several efforts have been carried out. In this paper, the angular flux discontinuity factor (AFDF) as the DF for the integro-differential transport equation (e.g., the discrete-ordinate method, the method of characteristics) is theoretically described and its applicability is discussed. The AFDF is used to preserve the region-wise neutron leakage at each spatial mesh and defined as a ratio of heterogeneous and homogeneous angular fluxes at the homogenized region surface. In a homogeneous calculation with the AFDF, the angular flux is discontinuous at the region surface. In this paper the applicability of the AFDF to fuel pin cell homogenization is verified for one-dimensional slab geometry. As a result of this verification, it is confirmed that the AFDF has the capability to reduce the spatial homogenization error of fuel pin cell homogenization.

DOI： 10.1080/00223131.2014.919882

Scopus

A new technique for spectral interference correction on pin-by-pin BWR core analysis 査読有り国際誌

Tatsuya Fujita, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 51 巻 ( 6 ) 頁： 783 - 797 2014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.

A new correction technique to capture the spectral interference effect on collapsed cross sections, which focuses on application to the pin-by-pin boiling water reactor (BWR) core analysis, is proposed. The spectral interference effect, which is caused by adjacent loadings of different types of fuel assemblies, has relationship with variations of neutron leakage in each pin-cell from the viewpoint of neutron balance. Variation of neutron leakage affects neutron spectrum and thus the neutron leakage is considered to be important to correct coarse-group cross sections used in core calculations. We focus on the neutron leakage in each pin-cell and use it as a correction index (i.e., a leakage index (LI)), which is defined as the volume-averaged neutron leakage in a pin-cell. By utilizing the leakage index, we represent the variations of coarse-group cross sections as the linear combination of LIs. In order to verify and discuss the applicability of the present correction technique, two-dimensional benchmark calculations considering typical characteristics of BWR cores are carried out. From the calculation results, the present correction technique well reproduces the reference coarse-group cross sections and improves the calculation accuracies.

DOI： 10.1080/00223131.2014.903212

Scopus

Cross section adjustment method based on random sampling technique 査読有り国際誌

Tomoaki Watanabe, Tomohiro Endo, Akio Yamamoto, Yasuhiro Kodama, Yasunori Ohoka, Tadashi Ushio

Journal of Nuclear Science and Technology 51 巻 ( 5 ) 頁： 590 - 599 2014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.

A cross section adjustment method based on the random sampling technique is proposed. In the proposed method, correlations among cross sections and core parameters are used instead of sensitivity coefficients of cross sections, which are necessary in the conventional method. The correlations are statistically estimated by the random sampling technique. The proposed method is theoretically consistent with the conventional method and provides comparable adjusted cross sections when sufficient number of random sampling is taken into account. The proposed method would be suitable for practical light water reactor (LWR) core analysis since estimation of sensitivity coefficients, which requires considerable computational cost in typical LWR problems, is not necessary. Through a benchmark problem in simple pin-cell geometry, adjusted cross sections by the present and the conventional cross section adjustment method are compared. The adjusted cross sections by the present method well reproduce the conventional ones, thus the feasibility of the present method is confirmed.

DOI： 10.1080/00223131.2014.882801

Scopus

A macroscopic cross-section model for BWR pin-by-pin core analysis 査読有り国際誌

Tatsuya Fujita, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 51 巻 ( 3 ) 頁： 282 - 304 2014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.

A macroscopic cross-section model used in boiling water reactor (BWR) pin-by-pin core analysis is studied. In the pin-by-pin core calculation method, pin-cell averaged cross sections are calculated for many combinations of core state and depletion history variables and are tabulated prior to core calculations. Variations of cross sections in a core simulator are caused by two different phenomena (i.e. instantaneous and history effects). We treat them through the core state variables and the exposure-averaged core state variables, respectively. Furthermore, the cross-term effect among the core state and the depletion history variables is considered. In order to confirm the calculation accuracy and discuss the treatment of the cross-term effect, the k-infinity and the pin-by-pin fission rate distributions in a single fuel assembly geometry are compared. Some cross-term effects could be negligible since the impacts of them are sufficiently small. However, the cross-term effects among the control rod history (or the void history) and other variables have large impacts
thus, the consideration of them is crucial. The present macroscopic cross-section model, which considers such dominant cross-term effects, well reproduces the reference results and can be a candidate in practical applications for BWR pin-by-pin core analysis on the normal operations.

DOI： 10.1080/00223131.2014.864248

Scopus

Application of augmented reality to nuclear reactor core simulation for fundamental nuclear engineering education 査読有り国際誌

Kosuke Tsujita, Tomohiro Endo, Akio Yamamoto

Nuclear Technology 185 巻 ( 1 ) 頁： 71 - 84 2014年1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 nuclear reactor core simulation system using augmented reality (AR) has been developed. Augmented reality is a technology that can provide additional information by overlaying computer graphics onto the image of actual world. In the past, AR has been applied to operation assistance in various fields. In the field of nuclear engineering, AR has been applied to support the decommissioning of nuclear power plants. Using AR, workers can simultaneously see the image of the actual world and helpful information
thus, they can intuitively imagine their works. This advantage of AR can be applied not only to operation assistance but also to other purposes. Therefore in this study we have tried to apply AR to a nuclear core simulator. The major purpose of the present AR core simulator is education of novice students who are not very familiar with nuclear reactor core behavior by enabling direct "hand manipulation" of a reactor model in the real world. For example users can directly insertlwithdraw a control rod and can directly see the power variation of the reactor through the AR technology.

DOI： 10.13182/NT13-7

Scopus

Uncertainty estimation of core safety parameters using cross-correlations of covariance matrix 査読有り国際誌

Akio Yamamoto, Yoshihiro Yasue, Tomohiro Endo, Yasuhiro Kodama, Yasunori Ohoka, Masahiro Tatsumi

Journal of Nuclear Science and Technology 50 巻 ( 10 ) 頁： 966 - 978 2013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：Taylor & Francis, Ltd.

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

DOI： 10.1080/00223131.2013.820155

Web of Science

Preservation of transmission probabilities in the method of characteristics 査読有り国際誌

Masato Tabuchi, Akio Yamamoto, Tomohiro Endo, Naoki Sugimura

Journal of Nuclear Science and Technology 50 巻 ( 8 ) 頁： 837 - 843 2013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.

An efficient calculation scheme for the method of characteristics, which allows coarser ray separation width, is newly proposed. Effective lengths of characteristics lines are evaluated from the viewpoint of preservation of transmission probability in the present method. The effective lengths to preserve transmission probabilities can be rigorously estimated through detailed numerical integrations. However, calculation and storage of all effective lengths is inefficient since the transmission probabilities depend on many parameters such as azimuthal angle, polar angle, position of characteristics line, and total cross section. To resolve this issue, in this paper, the analytical equations for the effective lengths are derived from an assumption that distribution function of actual characteristics lengths can be expressed by a linear function. The new approach introduced in this paper is named as transmission probability preservation through linear approximation (TPPL). Verification results indicate that accurate results can be obtained by the TPPL even if ray separation width becomes coarse.

DOI： 10.1080/00223131.2013.807753

Scopus

Convergence analysis of MOC inner iterations with large negative self-scattering cross-section 査読有り国際誌

Masato Tabuchi, Akio Yamamoto, Tomohiro Endo, Naoki Sugimura

Journal of Nuclear Science and Technology 50 巻 ( 5 ) 頁： 493 - 502 2013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.

When the transport correction is applied to the total cross-section, the self-scattering cross-section could have a negative value in order to preserve the balance of partial cross-sections. The negative self-scattering cross-section may lead to a negative impact on the convergence behavior for the method of characteristics (MOC), especially in a problem with large moderator regions containing hydrogen. In order to address this issue, the spectral radius of the inner iteration of MOC is theoretically estimated for various self-scattering cross-sections. It is found that the spectral radius of the inner iteration of MOC could exceed unity for a large negative self-scattering cross-section, which results in numerical divergence. A countermeasure for the divergence using the successive over relaxation method is also discussed in this paper.

DOI： 10.1080/00223131.2013.785271

Web of Science

Analysis of integral experiment on erbia-loaded thermal spectrum cores using Kyoto University Critical Assembly by MCNP code with various cross section libraries 査読有り国際誌

Yevgeniy Tur, Tomohiro Endo, Akio Yamamoto, Hironobu Unesaki, Masatoshi Yamasaki

Journal of Nuclear Science and Technology 49 巻 ( 10 ) 頁： 1028 - 1041 2012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：Taylor & Francis, Ltd.

Under the project on high burnup nuclear fuel development using erbium as a burnable poison, a series of experiments were performed at the Kyoto University Critical Assembly. The experimental results have formed the basis for this study which aims to analyze the suitability of various evaluated nuclear data libraries for using them in neutronic calculations under the project. The MCNP code was used for the analysis. Calculation model geometry was fully detailed, and ENDF, JENDL, JEFF, and TENDL libraries were used during calculation. For the cross sections of erbium nuclides, the analysis revealed that calculated results upon all the libraries corresponded with experimental data within the errors. However, in some libraries, significant differences were found in case of carbon and uranium nuclides under certain conditions.

DOI： 10.1080/00223131.2012.723174

Scopus

An optimization approach to establish an appropriate energy group structure for BWR pin-by-pin core analysis 査読有り国際誌

Tatsuya Fujita, Kenichi Tada, Tomohiro Endo, Akio Yamamoto, Shinya Kosaka, Go Hirano, Kenichiro Nozaki

Journal of Nuclear Science and Technology 49 巻 ( 7 ) 頁： 689 - 707 2012年7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 optimization approach to establish an appropriate multi-group energy structure for boiling water reactor (BWR) pin-by-pin finemesh core analysis is proposed. In the present approach, the number of energy groups of cross sections is successively reduced or increased. In order to select an energy group boundary that is removed or added, performances of all possible candidates of energy group structures are tested in multi-assembly geometries. Then, the energy group boundary, which provides theminimumdifference of the k-infinity or the pin-by-pin fission rate distribution, is finally removed or added. This procedure is repeated until the number of energy groups reaches to the target value. In order to confirm the applicability of the present approach, the accuracies of the k-infinity and the pin-by-pin fission rate distribution are investigated in various 2 × 2 multi-assembly geometries with the established energy group structure. From the verification results, the differences of the k-infinity and the pin-by-pin fission rate distribution between the reference (fine) and the established (coarse) energy group structure are small in the various 2 × 2 multiassembly geometries. Therefore, we can conclude that the present approach is efficient to establish an appropriate energy group structure for BWR pin-by-pin fine mesh core analysis.

DOI： 10.1080/00223131.2012.693890

Scopus

A unified approach for numerical calculation of space-dependent kinetic equation 査読有り国際誌

Yuichiro Ban, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 49 巻 ( 5 ) 頁： 496 - 515 2012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.

A unified numerical method based on the factorization approach is developed to solve the spacedependent neutron kinetic equation. Various numerical methods for solving the space-dependent kinetic equation have been developed so far. These methods can be classified into two categories, i.e., the direct and the factorization methods. The factorization method is known as an effective numerical method. In the present study, a new factorization method named the multigrid amplitude function (MAF) method is developed. Unlike the improved quasi-static (IQS) method, an independent amplitude function is assigned for each spatial region and energy group in the MAF method. The MAF method is a generalization of conventional methods, e.g., the frequency transform, IQS, and Theta methods. To evaluate the amplitude function in the MAF method, the time-dependent coarse-mesh finite difference (TCMFD) method is developed. The MAF method is implemented into a space-dependent kinetic code on the basis of the analytical polynomial nodal method. In order to verify the effectiveness of the MAF method, the TWIGL, Langenbuch, Maurer, and Werner (LMW), and Laboratorium fü r Reaktorregelung and Anlagensicherung (LRA) benchmark problems are analyzed. The calculation results show the effectiveness of the present method.

DOI： 10.1080/00223131.2012.677126

Scopus

Application of quick subchannel analysis method for three-dimensional pin-by-pin BWR core calculations 査読有り国際誌

Kenichi Tada, Tatsuya Fujita, Tomohiro Endo, Akio Yamamoto, Shinya Kosaka, Go Hirano, Kenichiro Nozaki

Journal of Nuclear Science and Technology 48 巻 ( 12 ) 頁： 1437 - 1452 2011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.

Three-dimensional pin-by-pin core analysis is considered to be a candidate for the next-generation BWR core calculation method. In our previous study, the applicability of the transport and burnup calculations for a three-dimensional pin-by-pin BWR core analysis was investigated. However, the thermal-hydraulics calculation has not yet been studied in this framework. In the conventional core analysis code, the bundlewise thermal-hydraulics calculation is adopted. In the actual core analysis, the power distribution inside a fuel assembly is tilted at the region adjacent to a control blade or the core peripheral region. In these regions, the consideration of the subchannel-wise void distribution has an impact on the fission rate distribution. Therefore, an evaluation of the detailed void distribution inside an assembly, i.e., the incorporation of the subchannel wise void distribution, is desirable for the pin-by-pin BWR core analysis. Although several subchannel analysis codes have been developed, these subchannel analysis codes generally require a large computational effort to estimate the subchannel-wise void distribution in a whole BWR core. Therefore, to analyze a whole BWR core within a reasonable computation time, it was necessary to apply a fast subchannel analysis code. In this paper, a quick subchannel analysis code dedicated to pin-by-pin BWR core analysis is newly developed, and the void distribution of the present subchannel analysis code is compared with the prevailing subchannel analysis code NASCA using three-dimensional single-assembly geometries. Since the present subchannel analysis code is used for a coupled neutronics/thermal-hydraulics analysis, the results of the coupling calculation are also compared with those of NASCA. The calculation result indicates that the void distribution difference between NASCA and the present subchannel analysis code is slightly less than 10%. This result indicates that the prediction accuracy of the present subchannel analysis code will be reasonably appropriate for a pin-by-pin BWR core analysis. Furthermore, the results show that the calculation time of the present subchannel analysis code is only 10 min for a hypothetical three-dimensional ABWR quarter-core geometry using a single CPU. This calculation time is sufficient for a pin-by-pin BWR core analysis.

DOI： 10.1080/18811248.2011.9711837

Scopus

Application of the robust design concept for fuel loading pattern 査読有り国際誌

Tomohiro Endo, Kazuma Ohori, Akio Yamamoto

Journal of Nuclear Science and Technology 48 巻 ( 7 ) 頁： 1077 - 1086 2011年7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.

Application of the robust design concept for fuel loading pattern design is proposed as a new approach to improve the prediction accuracy of core characteristics. The robust design is a design concept that establishes a resistant (robust) system for perturbations or noises, by properly setting design variables. In order to apply the concept of robust design to fuel loading pattern design, we focus on a theoretical approach based on the higher order perturbation method. This approach indicates that the eigenvalue separation is one of the effective indices to measure the robustness of a designed fuel loading pattern. In order to verify the effectiveness of the eigenvalue separation as an index of robustness, numerical analysis is carried out for typical 3-loop PWR cores, and we evaluated the correlation between the eigenvalue separation and the variation of relative assembly power due to the perturbation of the cross section. The numerical results show that the variation of relative power decreases as the eigenvalue separation increases
thus, it is confirmed that the eigenvalue separation is an effective index of robustness. Based on the eigenvalue separation of a fuel loading pattern, we discuss design guidelines of a fuel loading pattern to improve the robustness. For example, if each fuel assembly has independent uncertainty on its cross section, the robustness of the core can be enhanced by increasing the relative power at the center of the core. The proposed guidelines will be useful to design a loading pattern that has robustness for uncertainties due to cross section, calculation method, and so on.

DOI： 10.1080/18811248.2011.9711792

Scopus

Improved derivation of multigroup effective cross section for heterogeneous system by equivalence theory 査読有り国際誌

Akio Yamamoto, Tomohiro Endo, Hiroki Koike

Nuclear Science and Engineering 168 巻 ( 2 ) 頁： 75 - 92 2011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 validity of effective cross section obtained by the conventional equivalence theory is discussed from the viewpoint of reaction rate preservation in a heterogeneous system. It is shown that the reaction rate is not preserved when the escape probability is expressed by a multiterm rational approximation, which is commonly used in light water reactor (LWR) analyses. A new derivation method for obtaining a multigroup effective cross section, which accurately reproduces the result of reference ultrafine group calculation, is proposed. The validity of the proposed method is confirmed through test calculations in various heterogeneous geometries, which represent typical LWR configurations. Because the implementation of the proposed method is very simple, it is useful for existing lattice physics codes that utilize the equivalence theory on the basis of two-term (or multiterm) rational approximation.

DOI： 10.13182/NSE10-50

Scopus

Resonance calculation for large and complicated geometry using Tone's method by incorporating the method of characteristics 査読有り国際誌

Hao Yu, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology 48 巻 ( 3 ) 頁： 330 - 336 2011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.

A new efficient approach for evaluating the background cross section, which is based on Tone's method, is presented. Though the collision probability method is used in the conventional Tone's method, the method of characteristics (MOC) is used in the present method. Since the computation time of MOC is shorter than that of the collision probability method in a large and complicated geometry, the present method will be useful not only for lattice physics calculation, but also for analyses of advanced reactors with complicated geometry. Verification calculations are carried out in two configurations, i.e., a PWR fuel assembly geometry and a multiassembly geometry adjacent to the baffle-reflector region. The validity of the present method has been confirmed through the results of verification calculations.

DOI： 10.1080/18811248.2011.9711707

Scopus

Overview of Core Simulation Methodologies for Light Water Reactor Analysis 査読有り国際誌

A. Yamamoto, T. Endo

International Electronic Journal of Nuclear Safety and Simulation 2 巻 ( 1 ) 頁： 12 - 21 2011年3月

記述言語：英語掲載種別：研究論文（学術雑誌）

The current in-core fuel management calculation methods provide a very efficient route to predict neutronics behavior of light water reactor (LWR) cores and their prediction accuracy for current generation LWRs is generally sufficient. However, since neutronics calculations for LWRs are based on various assumptions and simplifications, we should also recognize many implicit limitations that are "embedded" in current neutronics calculation methodologies. Continuous effort for improvement of core simulation methodologies is also discussed.

Explicit Time Integration Scheme Using Krylov Subspace Method for Reactor Kinetics Equation 査読有り国際誌

Yuichiro Ban, Tomohiro Endo, Akio Yamamoto, Yoshihiro Yamane

Journal of Nuclear Science and Technology 48 巻 ( 2 ) 頁： 243 - 255 2011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 spatial discretization form of the space-dependent reactor kinetics equation is a first-order simultaneous ordinary differential equation in time. Conventional numerical methods of the space-dependent kinetics equation, i.e., the generalized Runge-Kutta method, the implicit method (backward Euler method), and the Theta method, are based on the time difference approximation. However, the present study adopts the analytical solution of the space-dependent kinetics equation expressed by the matrix exponential and no time difference approximation is used. In this context, our present approach is classified as an explicit method in which no iteration calculation on space and energy is necessary. The Krylov subspace method is used to evaluate the matrix exponential observed in the solution of the spatially discretized space-dependent kinetics equation. The Krylov subspace method is implemented into a space-dependent kinetics solver. In order to examine the effectiveness of the Krylov subspace method, the TWIGL benchmark problem is analyzed as a verification calculation. The calculation results show the effectiveness of the present method especially in the step reactivity perturbation.

DOI： 10.1080/18811248.2011.9711698

Scopus

Improvement of Tone's method with two-term rational approximation 査読有り国際誌

Akio Yamamoto, Tomohiro Endo, Go Chiba

Journal of Nuclear Science and Technology 48 巻 ( 2 ) 頁： 263 - 271 2011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.

An improvement of Tone's method, which is a resonance calculation method based on the equivalence theory, is proposed. In order to increase calculation accuracy, the two-term rational approximation is incorporated for the representation of neutron flux. Furthermore, some theoretical aspects of Tone's method, i.e., its inherent approximation and choice of adequate multigroup cross section for collision probability estimation, are also discussed. The validity of improved Tone's method is confirmed through a verification calculation in an irregular lattice geometry, which represents part of an LWR fuel assembly. The calculation result clarifies the validity of the present method.

DOI： 10.1080/18811248.2011.9711700

Scopus

AEGIS: An Advanced Lattice Physics Code for Light Water Reactor Analyses 査読有り国際誌

Akio Yamamoto, Tomohiro Endo, Masato Tabuchi, Naoki Sugimura, Tadashi Ushio, Masaaki Mori, Masahiro Tatsumi, Yasunori Ohoka

Nuclear Engineering and Technology 42 巻 ( 5 ) 頁： 500 - 519 2010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：Korean Nuclear Society

AEGIS is a lattice physics code incorporating the latest advances in lattice physics computation, innovative calculation models and efficient numerical algorithms and is mainly used for light water reactor analyses. Though the primary objective of the AEGIS code is the preparation of a cross section set for SCOPE2 that is a three-dimensional pin-by-pin core analysis code, the AEGIS code can handle not only a fuel assembly but also multi-assemblies and a whole core geometry in two-dimensional geometry. The present paper summarizes the major calculation models and part of the verification/validation efforts related to the AEGIS code.

DOI： 10.5516/NET.2010.42.5.500

Web of Science

その他リンク： https://doi.org/10.5516/NET.2010.42.5.500

Development of deterministic code based on the discrete ordinates method for the third-order neutron correlation technique 査読有り国際誌

Tomohiro Endo, Akio Yamamoto, Yoshihiro Yamane

Annals of Nuclear Energy 35 巻 ( 5 ) 頁： 927 - 936 2008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：Elsevier B.V.

The third-order neutron correlation technique is one of the subcriticality measurement techniques. The significant feature of this technique is that an absolute value of subcriticality can be evaluated from the measured neutron correlation factors, namely Y-infinity and X-infinity In order to extend the applicability of the third-order neutron correlation technique to actual complicated subcritical systems, we previously derived the spatial and neutron energy dependent theoretical formulas of Y-infinity and X-infinity by using the first-, second-, and third-order importance functions. These theoretical formulas depend on the locations of both an external neutron source and a neutron detector. The numerical solutions of these formulas are very useful to design the arrangements of the external neutron source and the neutron detector and to investigate the contamination of the spatial effect in the measured neutron correlation factors, Y-infinity and X-infinity Therefore, we developed a novel deterministic code that can calculate both Y-infinity and X-infinity based on the SN method. In this paper, we focus on the description of the numerical procedure of this code and its verification by comparing it with the results of Monte Carlo simulations.

DOI： 10.1016/j.anucene.2007.09.014

Web of Science

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