Faculty Profiles - ENDO Tomohiro

Papers - ENDO Tomohiro

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Benchmarks of subcriticality in accelerator-driven system at Kyoto University Critical Assembly Invited Reviewed International journal Open Access

Cheol Ho Pyeon, Masao Yamanaka, Song-Hyun Kim, Thanh-Mai Vu, Tomohiro Endo, Willem Fredrik G. Van Rooijen, Go Chiba

Nuclear Engineering and Technology Vol. 49 ( 6 ) page： 1234 - 1239 2017.9

Language：English Publishing type：Research paper (scientific journal) Publisher：Elsevier B.V.

Basic research on the accelerator-driven system is conducted by combining 235U-fueled and 232Th-loaded cores in the Kyoto University Critical Assembly with the pulsed neutron generator (14 MeV neutrons) and the proton beam accelerator (100 MeV protons with a heavy metal target). The results of experimental subcriticality are presented with a wide range of subcriticality level between near critical and 10,000 pcm, as obtained by the pulsed neutron source method, the Feynman-α method, and the neutron source multiplication method.

DOI： 10.1016/j.net.2017.06.012

Recent developments in the GENESIS code based on the Legendre polynomial expansion of angular flux method Invited Reviewed International journal Open Access

Akio Yamamoto, Akinori Giho, Tomohiro Endo

Nuclear Engineering and Technology Vol. 49 ( 6 ) page： 1143 - 1156 2017.9

Language：English Publishing type：Research paper (scientific journal) Publisher：Elsevier B.V.

This 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, the concept of the simplified Pn method is introduced in order to reduce storage of flux moments. The accuracy of the present method is verified through a benchmark problem. Next, the iteration stability of the GENESIS code for the highly voided condition, which would appear in a severe accident (e.g., design extension) conditions, is discussed. The efficiencies of the coarse mesh finite difference and generalized coarse mesh rebalance acceleration methods are verified with various stabilization techniques. Use of the effective diffusion coefficient and the artificial grid diffusion coefficients are found to be effective to stabilize the acceleration calculation in highly voided conditions.

DOI： 10.1016/j.net.2017.06.016

Experimental benchmarks on kinetic parameters in accelerator-driven system with 100 MeV protons at Kyoto University Critical Assembly Reviewed International journal

Cheol Ho Pyeon, Masao Yamanaka, Tomohiro Endo, Willem Fredrik G. van Rooijen, Go Chiba

Annals of Nuclear Energy Vol. 105 page： 346 - 354 2017.7

Language：English Publishing type：Research paper (scientific journal) Publisher：Elsevier B.V.

Accelerator-driven system experiments with spallation neutrons (100 MeV protons and Pb-Bi target) are carried out in the 235U-fueled and Pb-Bi-zoned core at the Kyoto University Critical Assembly, under a subcritical state ranging between 1160 and 11,556 pcm. In these experiments, measurement of the prompt neutron decay constant and the subcriticality is conducted by the pulsed neutron source (PNS) method and the Feynman-α method with the use of optical fiber detectors. The experimental results successfully validate the prompt neutron decay constant and the subcriticality through the deduction of kinetic parameters by both the PNS and the α-fitting methods. The detector position dependency, neutron spectrum and subcriticality measurement methods still remain, however, in these experiments. For onward studies, the experimental benchmarks obtained from these experiments are expected to be involved in the numerical verification of subcriticality on-line monitoring, in the analysis of subcriticality uncertainty and in the deterministic approach to kinetic parameters.

DOI： 10.1016/j.anucene.2017.03.030

Estimation of sensitivity coefficients of core characteristics based on reduced-order modeling using sensitivity matrix of assembly characteristics Reviewed International coauthorship International journal Open Access

Ryota Katano, Tomohiro Endo, Akio Yamamoto, Mohammad Abdo, Hany Abdel-Khalik

Journal of Nuclear Science and Technology Vol. 54 ( 6 ) page： 637 - 647 2017.6

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

We propose an estimation method of sensitivity coefficients of core neutronics parameters based on a multi-level reduced-order modeling approach. The idea is to use lower-level models to identify the dominant input parameter variations, constrained to the so-called active subspace, which are employed to determine the sensitivity coefficients of the core neutronic parameters. In our implementation, the lower-level model is represented by two-dimensional assembly calculations, which are employed in the preparation of the few-group cross-sections for core-wide calculations. The active subspace basis is estimated using the singular value decomposition of sensitivity matrices of assembly neutronics parameters. In numerical verification calculation, sensitivity coefficients of core characteristics for a typical three-loop PWR equilibrium-cycle are estimated using the proposed method and the direct method. Comparison of these two results shows that the proposed method well reproduces the results obtained by the direct method with lower calculation costs. Through the verification calculations, applicability of the proposed method to practical light water reactor analysis is confirmed.

DOI： 10.1080/00223131.2017.1299052

Estimation of modeling approximation errors using data assimilation with the minimum variance approach Reviewed International journal

Akio Yamamoto, Kuniharu Kinoshita, Tomohiro Endo

Journal of Nuclear Science and Technology Vol. 54 ( 4 ) page： 459 - 471 2017.4

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

In this paper, we estimate prediction errors owing to approximations in calculation models (modeling approximation error) using the data assimilation method. Correlations between the modeling approximation error and neutronics parameters obtained through calculations are evaluated in test configurations and then the evaluated correlations are used to predict the modeling approximation errors in design configuration. Formulae to estimate the modeling approximation error using the correlations are derived based on the minimum variance approach and the physical interpretation of the formulae is discussed through simple cases. The proposed method is applied in 2 × 2 and 3 × 3 fuel assembly geometries using specifications of the KAIST benchmark problem. The correlation between the modeling approximation error and parameters (neutron leakage in each fuel assembly) is estimated in 2 × 2 fuel assemblies and then the modeling approximation error in 3 × 3 fuel assemblies is predicted using the correlation. The calculation results not only indicate feasibility of the present method, but also suggest a need for further investigation on the assumptions used in the present study, i.e. applicability and robustness of the correlation among different geometries.

DOI： 10.1080/00223131.2017.1286271

GENESIS: A three-dimensional heterogeneous transport solver based on the legendre polynomial expansion of angular flux method Reviewed International journal

Akio Yamamoto, Akinori Giho, Yuki Kato, Tomohiro Endo

Nuclear Science and Engineering Vol. 186 ( 1 ) page： 1 - 22 2017.4

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

A heterogeneous transport solver in three-dimensional (3-D) geometry, GENESIS, is developed incorporating recent developments in the method of characteristics (MOC) in 3-D geometry. The Legendre Polynomial Expansion of Angular Flux (LEAF) method is implemented in the GENESIS code, in which neutron transport is calculated in two-dimensional (2-D) characteristics planes rather than in one-dimensional characteristics lines adopted in the conventional approach of 3-D MOC. Unlike the planar MOC method that combines 2-D 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 planarMOC method: No approximation is used for axial leakage treatment. The GENESIS code can handle flexible shapes of objects in rectangular or hexagonal geometry. A two-level, multigroup generalized coarse mesh rebalance 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 based on the LEAF method.

DOI： 10.1080/00295639.2016.1273002

Automated generation of burnup chain for reactor analysis applications Reviewed International coauthorship International journal

Viet-Phu Tran, Hoai-Nam Tran, Akio Yamamoto, Tomohiro Endo

Kerntechnik Vol. 82 ( 2 ) page： 196 - 205 2017.4

Language：English Publishing type：Research paper (scientific journal) Publisher：Carl Hanser Verlag

This paper presents the development of an automated generation of burnup chain for reactor analysis applications. Algorithms are proposed to reevaluate decay modes, branching ratios and effective fission product (FP) cumulative yields of a given list of important FPs taking into account intermediate reactions. A new burnup chain is generated using the updated data sources taken from the JENDL FP decay data file 2011 and Fission yields data file 2011. The new burnup chain is output according to the format for the SRAC code system. Verification has been performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Burnup calculations using the new burnup chain have also been performed based on UO2 and MOX fuel pin cells and compared with a reference chain th2cm6fp193bp6T.

DOI： 10.3139/124.110671

An efficient execution of Monte Carlo simulation based on delta-tracking method using GPUs Reviewed International journal

Takuya Okubo, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology Vol. 54 ( 1 ) page： 30 - 38 2017.1

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

An efficient execution method for Monte Carlo simulation using graphic processing unit (GPU) is proposed. The delta-tracking method is used since the delta-tracking method can reduce conditional branches and complexity of code implementation, which degrade computational performance on GPUs. In order to improve parallel efficiency in the eigenvalue calculation, generated fission neutrons are recorded using the atomic operation which avoids the data race in GPUs. We also propose a method to efficiently tally neutron flux in a region. The present method is based on the atomic operation and use of fixed-point type number instead of common floating-point type number. The verification calculations using the C5G7 benchmark problem show effectiveness of the proposed numerical algorithms on GPUs through comparison with calculations using central processing units.

DOI： 10.1080/00223131.2016.1202793

Reduction of MOC Discretization Errors through a Minimization of Source Ratio Variances Reviewed International journal Open Access

M. Tabuchi, A. Yamamoto, T. Endo, M. Tatsumi

Journal of Nuclear Science and Technology Vol. 53 ( 11 ) page： 1858 - 1869 2016.11

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

A new technique to reduce discretization errors for ray tracing in the method of characteristics (MOC) is proposed focusing on depletion calculations of single and multi-assembly geometries. In order to efficiently carry out depletion calculations, a calculation scheme using the superhomogenization (SPH) method can be used. However, the discretization errors are caused by changes of neutron sources and total cross sections according to a depletion. This fact means that improvement of accuracy cannot be expected by the calculation scheme with the SPH method when changes of the above parameters are significant. In order to mitigate this problem, a new approach is developed. In the new approach, the discretization errors are reduced by minimizing a variance of a certain parameter which is composed of a ratio of neutron source to total cross section. The verification results suggest that accuracy is degraded by the SPH method as expected especially in a geometry where neutron sources and total cross sections are drastically changing through a depletion. On the other hand, the new approach gives more accurate results compared to the conventional MOC in all calculation cases. Consequently, improvement of calculation efficiency by the new approach is confirmed.

DOI： 10.1080/00223131.2016.1171172

Research data storage URL： http://hdl.handle.net/2237/25272

Nuclear data-induced uncertainty quantification of neutronics parameters of accelerator-driven system Reviewed International journal Open Access

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

Journal of Nuclear Science and Technology Vol. 53 ( 10 ) page： 1653 - 1661 2016.10

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

Nuclear data-induced uncertainties of neutronics parameters of one accelerator-driven system concept designed by the Japan Atomic Energy Agency are quantified. The variance-covariance data provided in the JENDL-4.0 library are used. Uncertainties are quantified for effective neutron multiplication factor, subcritical neutron multiplication rate, a family of delayed neutron fractions, power peaking and coolant void reactivity at several operational states. Inter-cycle and inter-parameter correlation matrices and detailed information such as nuclide-wise and nuclear data-wise uncertainties are also provided.

DOI： 10.1080/00223131.2015.1127185

A CMFD acceleration method for SP3 advanced nodal method Reviewed International journal

Akio Yamamoto, Tatsuya Sakamoto, Tomohiro Endo

Nuclear Science and Engineering Vol. 184 ( 2 ) page： 168 - 173 2016.10

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

Flux-level-fixup (FF) coarse-mesh finite difference (CMFD) (FF-CMFD), which increases numerical stability during nonlinear iterations for the SP3 advanced nodal method, is proposed as an improved CMFD implementation. In contrast to the scalar flux that appeared in the advanced nodal method with diffusion theory, the second flux moment φ2 in the SP3 method could take a very small value since it can take both positive and negative values in a node. This is a root cause of inefficiency of the SP3 advanced nodal method when conventional CMFD acceleration is directly applied. In the proposed FF-CMFD method, a constant value is added to the second flux moment φ2 to fix up its value to a sufficiently large positive value for stable numerical calculations. The efficiency of the FF-CMFD method is verified through benchmark calculations.

DOI： 10.13182/NSE16-53

Statistical error estimation of the Feynman-α method using the bootstrap method Reviewed International journal Open Access

Tomohiro Endo, Akio Yamamoto, Takahiro Yagi, Cheol Ho Pyeon

Journal of Nuclear Science and Technology Vol. 53 ( 9 ) page： 1447 - 1453 2016.9

Authorship：Lead author,　Corresponding author Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

Applicability of the bootstrap method is investigated to estimate the statistical error of the Feynman-α method, which is one of the subcritical measurement techniques on the basis of reactor noise analysis. In the Feynman-α method, the statistical error can be simply estimated from multiple measurements of reactor noise, however it requires additional measurement time to repeat the multiple times of measurements. Using a resampling technique called “bootstrap method,” standard deviation and confidence interval of measurement results obtained by the Feynman-α method can be estimated as the statistical error, using only a single measurement of reactor noise. In order to validate our proposed technique, we carried out a passive measurement of reactor noise without any external source, i.e. with only inherent neutron source by spontaneous fission and (α,n) reactions in nuclear fuels at the Kyoto University Criticality Assembly. Through the actual measurement, it is confirmed that the bootstrap method is applicable to approximately estimate the statistical error of measurement results obtained by the Feynman-α method.

DOI： 10.1080/00223131.2015.1113898

Research data storage URL： http://hdl.handle.net/2237/25270

Discontinuity factors for simplified P3 theory Reviewed International journal

Akio Yamamoto, Tatsuya Sakamoto, Tomohiro Endo

Nuclear Science and Engineering Vol. 183 ( 1 ) page： 39 - 51 2016.5

Language：English Publishing type：Research paper (scientific journal) Publisher：American Nuclear Society

New discontinuity factors (DFs), i.e., individual and common DFs, for the simplified P3 (SP3) theory are proposed. In the individual DFs, two DFs are used for zeroth- and second-order angular moments in order to preserve first- and third-order angular moments of SP3 at a surface of the homogenized region. Contrarily, the same value of DF is used for zeroth- and second-order angular moments, and the first-order angular moment is preserved in the common DF. Theoretical derivation for these DFs are described, and then, actual numerical calculation procedures for these DFs are explained. Verification results in color-set geometries loaded with UO2 and mixed oxide fuel assemblies indicate the validity of the present method for cell-homogenized pin-by-pin SP3 calculations. Homogenization errors on keff and pin-power distribution are significantly reduced by the present DFs. The proposed DFs can be used for practical pin-by-pin core analyses using the SP3 theory.

DOI： 10.13182/NSE15-102

Uncertainty Quantification of LWR Core Characteristics Using Random Sampling Method Reviewed International journal

Akio Yamamoto, Kuniharu Kinoshita, Tomoaki Watanabe, Tomohiro Endo, Yasuhiro Kodama, Yasunori Ohoka, Tadashi Ushio, Hiroaki Nagano

Nuclear Science and Engineering Vol. 181 ( 2 ) page： 160 - 174 2015.10

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

Uncertainties of various neutronics characteristics in commercial boiling water reactor (BWR) and pressurized water reactor (PWR) cores due to cross-section covariance are evaluated by the Latin Hypercube Sampling (LHS) method, which is an efficient random sampling algorithm. Thermal-hydraulic feedback and burnup effects are fully and explicitly taken into account using a licensing-grade core simulator. Uncertainties for various core characteristics are evaluated by the statistical processing of core calculation results based on the LHS method. The calculation results indicate that uncertainty of critical eigenvalue (i.e., core reactivity) in the BWR core is comparable to that of a typical PWR core. On the other hand, uncertainties of assembly relative power distribution and maximum assembly burnup in the present BWR core are much smaller than those of the present PWR core. The strong thermal-hydraulic feedback effect in the BWR core significantly contributes to the difference of uncertainties in BWR and PWR cores.

DOI： 10.13182/NSE14-152

Application of correction technique using leakage index combined with SPH or discontinuity factors for energy collapsing on pin-by-pin BWR core analysis Reviewed International journal Open Access

Tatsuya Fujita, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology Vol. 52 ( 3 ) page： 355 - 370 2015.3

Language：English Publishing type：Research paper (scientific journal) Publisher：Taylor & Francis, Ltd.

A correction technique to capture the spectral interference effect on collapsed cross sections is combined with the superhomogenization (SPH) factor or the discontinuity factor (DF) and is applied to the pin-by-pin core analysis for boiling water reactors (BWRs). The spectral interference effect has relationship with variations of neutron leakage in each pin-cell from the viewpoint of neutron balance. In order to correct collapsed cross sections, a new correction technique, in which the neutron leakage in each pin-cell is used as a correction index, was proposed in the previous study. By this correction technique, the reference coarse group cross sections are well reproduced and the calculation accuracies are improved. However, the reference fine group calculation results could not be reproduced since the correction technique cannot reduce energy collapsing errors. Thus, we combine the correction technique with the SPH factor or the DF to reduce energy collapsing errors. In order to verify and discuss the applicability of the correction technique with the SPH factor or the DF, two-dimensional benchmark calculations considering typical characteristics of BWR cores are carried out. The correction technique with the DF more accurately reproduces the reference fine group calculation results than that with the SPH factor.

DOI： 10.1080/00223131.2014.948520

Applicability of angular flux discontinuity factor preserving region-wise leakage for integro-differential transport equation Reviewed International journal Open Access

Tatsuya Sakamoto, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology Vol. 51 ( 10 ) page： 1264 - 1273 2014.10

Language：English Publishing type：Research paper (scientific journal) Publisher：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

A new technique for spectral interference correction on pin-by-pin BWR core analysis Reviewed International journal Open Access

Tatsuya Fujita, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology Vol. 51 ( 6 ) page： 783 - 797 2014.6

Language：English Publishing type：Research paper (scientific journal) Publisher：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

Cross section adjustment method based on random sampling technique Reviewed International journal

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

Journal of Nuclear Science and Technology Vol. 51 ( 5 ) page： 590 - 599 2014.5

Language：English Publishing type：Research paper (scientific journal) Publisher：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

A macroscopic cross-section model for BWR pin-by-pin core analysis Reviewed International journal

Tatsuya Fujita, Tomohiro Endo, Akio Yamamoto

Journal of Nuclear Science and Technology Vol. 51 ( 3 ) page： 282 - 304 2014.3

Language：English Publishing type：Research paper (scientific journal) Publisher：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

Application of augmented reality to nuclear reactor core simulation for fundamental nuclear engineering education Reviewed International journal

Kosuke Tsujita, Tomohiro Endo, Akio Yamamoto

Nuclear Technology Vol. 185 ( 1 ) page： 71 - 84 2014.1

Language：English Publishing type：Research paper (scientific journal) Publisher：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

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