Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Nagaya, Yasunobu
EPJ Nuclear Sciences & Technologies (Internet), 11, p.1_1 - 1_7, 2025/01
Japan Atomic Energy Agency (JAEA) has been developing a general-purpose continuous-energy Monte Carlo code MVP for nuclear reactor core analysis. Recently improvements to MVP have been focused on the development of an advanced neutronics/thermal-hydraulics coupling code. JAEA has also developed a new Monte Carlo solver Solomon for criticality safety analysis. Solomon aims to calculate the criticality of a damaged reactor core including fuel debris. This paper provides an overview of the capabilities and reviews recent applications of MVP and Solomon.
Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu
Nuclear Science and Engineering, 199(1), p.18 - 41, 2025/01
Times Cited Count：1 Percentile：62.55(Nuclear Science & Technology)A series of integral experiments were conducted at FCA of JAEA, simulating LWR cores with a tight lattice cell of highly enriched MOX fuel containing more than 15% fissile plutonium. The three experimental configurations were constructed using foamed polystyrene with different void fractions to clarify the prediction accuracy of neutronic calculation codes and nuclear data among various neutron spectra. The nuclear characteristics measured in the experiments were criticality, moderator void reactivity worths, and sample reactivity worths. The preliminary analyses on experiments were conducted using a deterministic calculation code conventionally used for fast reactors with JENDL-4.0. Most reactivity worth calculations correlated well with the experimental values. Specifically for the softer neutron spectra configurations, the treatment of ultrafine energy groups obviously improved the prediction accuracy of the deterministic calculations. Furthermore, reference calculations were performed with MVP3 code by modeling the experimental setup in detail, confirming the validity of the deterministic calculations.
Kamiya, Tomohiro; Nagatake, Taku; Ono, Ayako; Tada, Kenichi; Kondo, Ryoichi; Nagaya, Yasunobu; Yoshida, Hiroyuki
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 7 Pages, 2024/11
We have developed the JAEA Advances Multi-Physics Analysis platform for Nuclear systems (JAMPAN) to realize high-fidelity neutronics/thermal-hydraulics coupling simulations. We will perform MVP/JUPITER coupling simulation for a single BWR fuel assembly in order to confirm that the neutronics/thermal-hydraulics coupling through JAMPAN is feasible. This presentation explains how to send and receive data between MVP and JUPITER through JAMPAN and simulation results.
Tuya, D.; Nagaya, Yasunobu
Nuclear Science and Engineering, 198(5), p.1021 - 1035, 2024/05
Times Cited Count：1 Percentile：27.70(Nuclear Science & Technology)In Monte Carlo neutron transport calculations for local response or deep penetration problems, some estimation of an importance function is generally required in order to improve their efficiency. In this work, a new recursive Monte Carlo (RMC) method, which is partly based on the original RMC method, for estimating an importance function for local variance reduction (i.e., source-detector type) problems has been developed. The new RMC method has been applied to two sample problems of varying degrees of neutron penetrations, namely a one-dimensional iron slab problem and a three-dimensional concrete-air problem. The biased Monte Carlo calculations with variance reduction parameters based on the obtained importance functions by the new RMC method have been performed to estimate detector responses in these problems. The obtained results are in agreement with those by the reference unbiased Monte Carlo calculations. Furthermore, the biased calculations offered an increase in efficiency on the order of 1 to 10
in terms of the figure of merit (FOM). The results also indicated that the efficiency increased as the neutron penetration became deeper.
Tada, Kenichi; Kondo, Ryoichi; Kamiya, Tomohiro; Nagatake, Taku; Ono, Ayako; Nagaya, Yasunobu; Yoshida, Hiroyuki
Proceedings of International Conference on Physics of Reactors (PHYSOR 2024) (Internet), p.1488 - 1497, 2024/04
JAEA has developed a new high-fidelity multi-physics platform JAMPAN for connecting single-physics codes such as a neutronics code and a thermal-hydraulics code. It consists of the HDF5 formatted data container and input and output data handler modules to generate the input file and read the output file of the single-physics code. Users can easily add or exchange the code by implementing input and output data handler modules for this code. The first target of JAMPAN is the coupling of neutronics and thermal-hydraulics calculations to provide reference results of core analysis codes. The current version of JAMPAN couples the neutronics code MVP and the thermal-hydraulics codes JUPITER, ACE-3D, and NASCA. Users can select the thermal-hydraulics code depending on the scale of problems to be solved, computational performance, and so on. This presentation explains the overview of JAMPAN and shows the results of the neutronics and thermal-hydraulics coupling calculation.
Suzuki, Motomu*; Nagaya, Yasunobu
Journal of Nuclear Science and Technology, 61(2), p.177 - 191, 2024/02
Times Cited Count：1 Percentile：0.00(Nuclear Science & Technology)With the release of the latest Japanese evaluated nuclear data library JENDL-5, the prediction accuracy of JENDL-5 for neutronics parameters of the BEAVRS benchmark for the hot zero power condition was evaluated in this study. The criticality, control rod bank worth (CRW), isothermal temperature coefficient (ITC), and in-core detector signals were calculated and compared with the measured data for evaluation. For the criticality, the calculation-to-experimental (C/E) values varied between 1.0001 and 1.0045. Sensitivity analysis by replacing cross section data from the JENDL-4.0u1 with JENDL-5 revealed that 
H, 
U, 
U, and 
O significantly affected the criticality. The individual CRW agreed within 50 pcm, and total CRW also agreed within 100 pcm from the measured values. The ITC results calculated with a temperature deviation of 5.56 K case were negatively overestimated comparing with the measured values; whereas those of with 2.78 K were improved and agree with the measured values within a standard deviation. The axial detector signals indicated a maximum relative error of 4.46% and the root mean squared error (RMSE) of 2.13%. The differences between the previous version of JENDL-4.0u1 and JENDL-5 were also investigated.
Tada, Kenichi; Nagaya, Yasunobu; Taninaka, Hiroshi; Yokoyama, Kenji; Okita, Shoichiro; Oizumi, Akito; Fukushima, Masahiro; Nakayama, Shinsuke
Journal of Nuclear Science and Technology, 61(1), p.2 - 22, 2024/01
Times Cited Count：11 Percentile：96.03(Nuclear Science & Technology)The new version of the Japanese evaluated nuclear data library, JENDL-5, was released in December 2021. This paper demonstrates the validation of JENDL-5 for fission reactor applications. Benchmark calculations are performed with the continuous-energy Monte Carlo codes MVP and MCNP and the deterministic code system MARBLE. The benchmark calculation results indicate that the performance of JENDL-5 for fission reactor applications is better than that of the former library JENDL-4.0.
Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu
Nuclear Science and Engineering, 15 Pages, 2024/00
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Tuya, D.; Nagaya, Yasunobu
Journal of Nuclear Engineering (Internet), 4(4), p.691 - 710, 2023/11
The Monte Carlo method is used to accurately estimate various quantities such as k-eigenvalue and integral neutron flux. However, when a distribution of a quantity is desired, the Monte Carlo method does not typically provide continuous distribution. Recently, the functional expansion tally and kernel density estimation methods have been developed to provide continuous distribution. In this paper, we propose a method to estimate a continuous distribution of a quantity using artificial neural network (ANN) model with Monte Carlo-based training data. As a proof of concept, a continuous distribution of iterated fission probability (IFP) is estimated by ANN models in two systems. The IFP distributions by the ANN models were compared with the Monte Carlo-based data and the adjoint angular neutron fluxes by the PARTISN code. The comparisons showed varying degrees of agreement or discrepancy; however, it was observed that the ANN models learned the general trend of the IFP distributions.
Suyama, Kenya; Ueki, Taro; Gunji, Satoshi; Watanabe, Tomoaki; Araki, Shohei; Fukuda, Kodai; Yamane, Yuichi; Izawa, Kazuhiko; Nagaya, Yasunobu; Kikuchi, Takeo; et al.
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 6 Pages, 2023/10
To remove and store safely the fuel debris generated by the severe accident of the Fukushima Daiichi Nuclear Power Station in 2011 is one of the most important and challenging topics for decommissioning of the damaged reactors in Fukushima. To validate the adopted method for the evaluation of criticality safety control of the fuel debris through comparison with the experimental data obtained by the criticality experiments, the Nuclear Regulation Authority (NRA) of Japan funds a research and development project which was entrusted to the Nuclear Safety Research Center (NSRC) of Japan Atomic Energy Agency (JAEA) from 2014. In this project, JAEA has been conducting such activities as i) comprehensive computation of the criticality characteristics of the fuel debris and making database (criticality map of the fuel debris), ii) development of new continuous energy Monte Carlo code, iii) evaluation of criticality accident and iv) modification of the critical assembly STACY for the experiments for validation of criticality safety control methodology. After the last ICNC2019, the project has the substantial progress in the modification of STACY which will start officially operation from May 2024 and the development of the Monte Carlo Code "Solomon" suitable for the criticality calculation for materials having spatially random distribution complies with the power spectrum. We present the whole picture of this research and development project and status of each technical topics in the session.
Kondo, Ryoichi; Nagaya, Yasunobu
Proceedings of International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2023) (Internet), 10 Pages, 2023/08
A functional expansion tally (FET) method with numerical basis functions generated by singular value decomposition (SVD) is newly proposed. Traditionally, analytical functions were used for the FET calculations, e.g., Legendre polynomials for a one-dimensional distribution. However, the expansion terms could increase to reconstruct steep or complex distributions with these functions. A basis set that can well represent the target distribution with lower order expansion is desired to achieve high accuracy with the small computational resource. In the present study, a numerical basis set is generated from snapshot data by using SVD. This approach is based on the reduced order modeling (ROM). We applied ROM to the FET method in Monte Carlo calculations. The numerical result showed the applicability of the proposed method, on the other hand, some issues were revealed, e.g., discretization of the snapshot data.
Iwamoto, Osamu; Iwamoto, Nobuyuki; Kunieda, Satoshi; Minato, Futoshi; Nakayama, Shinsuke; Kimura, Atsushi; Nakamura, Shoji; Endo, Shunsuke; Nagaya, Yasunobu; Tada, Kenichi; et al.
EPJ Web of Conferences, 284, p.14001_1 - 14001_7, 2023/05
Times Cited Count：2 Percentile：87.25(Nuclear Science & Technology)Tuya, D.; Nagaya, Yasunobu
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05
Adjoint angular neutron flux is used as weighting function in various reactor applications such as calculation of kinetics parameters, importance sampling variance reduction techniques, etc. Iterated fission probability (IFP), which is proportional to a fundamental mode of adjoint angular neutron flux, has increasingly been used as weighting function in Monte Carlo calculations. The Monte Carlo based IFP methods stochastically estimate IFP for a given phase-space location. In this work, we investigated the applicability of a deep neural network for approximating an unknown underlying function, which maps from a phase-space location to an IFP in a given fissile system, from dataset produced by a Monte Carlo based IFP method. The preliminary application has been performed for the Godiva and simplified STACY cores. The comparison showed a varying degree of agreement and discrepancy between the results by the DNN and the reference results by a deterministic neutron transport code PARTISN.
Tada, Kenichi; Yamamoto, Akio*; Kunieda, Satoshi; Nagaya, Yasunobu
JAEA-Data/Code 2022-009, 208 Pages, 2023/02
JAEA-Data-Code-2022-009.pdf:3.87MB
The nuclear data processing code has an important role to connect evaluated nuclear data libraries and neutronics calculation codes. Japan Atomic Energy Agency (JAEA) has developed the nuclear data processing code FRENDY since 2013 to generate cross section files from evaluated nuclear data libraries, such as JENDL, ENDF/B, JEFF, and TENDL. The first version of FRENDY was released in 2019. FRENDY version 1 generates ACE files which are used for continuous energy Monte Carlo codes such as PHITS, Serpent, and MCNP. FRENDY version 2 generates multi-group neutron cross-section files from ACE files. The other major improvements are as follows: (1) uncertainty quantification for the probability tables of the unresolved resonance cross-section; (2) perturbation of the ACE file for the uncertainty quantification using a continuous Monte Carlo code; (3) modification of the ENDF-6 formatted nuclear data file. This report describes an overview of the nuclear data processing methods and input instructions for FRENDY.
Iwamoto, Osamu; Iwamoto, Nobuyuki; Kunieda, Satoshi; Minato, Futoshi; Nakayama, Shinsuke; Abe, Yutaka*; Tsubakihara, Kosuke*; Okumura, Shin*; Ishizuka, Chikako*; Yoshida, Tadashi*; et al.
Journal of Nuclear Science and Technology, 60(1), p.1 - 60, 2023/01
Times Cited Count：187 Percentile：99.98(Nuclear Science & Technology)Kamiya, Tomohiro; Ono, Ayako; Tada, Kenichi; Akie, Hiroshi; Nagaya, Yasunobu; Yoshida, Hiroyuki; Kawanishi, Tomohiro
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 8 Pages, 2022/11
JAEA started to develop the advanced reactor analysis code JAMPAN (JAEA advanced multi-physics analysis platform for nuclear systems). The current version of JAMPAN handles the continuous energy Monte Carlo code MVP and the detailed thermal-hydraulics analysis code for multiphase and multicomponent JUPITER. JAMPAN is designed to consider the extensibility and it does not depend on the analysis codes. All calculations in JAMAPAN are not directly connected. JAMPAN has data containers, and all input and output data of each analysis code are set in these data containers. JAMPAN will easily exchange the calculation codes and add the other calculations, e.g., structure calculation and irradiation calculation since the input and the output format of each code has no impact on the other calculation codes. The 4 by 4 pin-cell geometry was used as the demonstration calculation of JAMPAN and the physically reasonable calculation results were obtained.
-eigenvalue perturbation in Monte Carlo calculationTuya, D.; Nagaya, Yasunobu
Annals of Nuclear Energy, 169, p.108919_1 - 108919_9, 2022/05
Times Cited Count：4 Percentile：54.60(Nuclear Science & Technology)Estimating an effect of a perturbation in a fissile system on its -eigenvalue requires special technique called perturbation theory when the considered perturbation is small. In this study, we develop an adjoint-weighted correlated sampling (AWCS) method based on the exact perturbation theory without any approximation by combining the correlated sampling (CS) method with iterated-fission probability (IFP) based adjoint-weighting method. With the advantages of the CS method being good at providing very small uncertainty for small perturbations and the IFP-based adjoint-weighting method being suitable for continuous-energy Monte Carlo calculation, the developed AWCS method based on the exact perturbation theory offers a new rigorous approach for perturbation calculations. The obtained results by the developed AWCS method for verification problems involving Godiva and simplified STACY density perturbations showed good agreement with the reference calculations.
Okita, Shoichiro; Nagaya, Yasunobu; Fukaya, Yuji
Journal of Nuclear Science and Technology, 58(9), p.992 - 998, 2021/09
Times Cited Count：2 Percentile：21.34(Nuclear Science & Technology)Tada, Kenichi; Kunieda, Satoshi; Nagaya, Yasunobu
Proceedings of 2019 IEEE Nuclear Science Symposium and Medical Imaging Conference (NSS/MIC 2019), Vol.1, p.48 - 49, 2020/08
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Nuclear data processing is an important interface between evaluated nuclear data library and neutronics transport codes. JAEA developed a new nuclear data processing code FRENDY in order to process the evaluated nuclear data library JEND. JAEA released FRENDY version 1 as an open source software under the 2-clause BSD license. The current version of FRENDY can generate ACE formatted files for continuous energy Monte Carlo calculation codes such as MCNP and PHITS. It uses the same processing method as NJOY because the implementation of the conventional method is an important step to develop the new code. For verification, we compared the processing results of FRENDY with those of NJOY and confirmed FRENDY provided us with almost the same outputs as NJOY. In this presentation, we explains the overview and a future plan of FRENDY.
Nagaya, Yasunobu; Ueki, Taro; Tonoike, Kotaro
Proceedings of 11th International Conference on Nuclear Criticality Safety (ICNC 2019) (Internet), 9 Pages, 2019/09
A new Monte Carlo solver Solomon has been developed for the application to fuel-debris systems. It is designed not only for usual criticality safety analysis but also for criticality calculations of damaged reactor core including fuel debris. This paper describes the current status of Solomon and demonstrates the applications of the randomized Weierstrass function (RWF) model and the RWF model superimposed voxel geometry.
