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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.
Hashimoto, Shintaro; Sato, Tatsuhiko
Journal of Nuclear Science and Technology, 56(4), p.345 - 354, 2019/04
Times Cited Count：5 Percentile：47.59(Nuclear Science & Technology)Particle transport simulations based on the Monte Carlo method have been applied to shielding calculations. Estimation of not only statistical uncertainty related to the number of trials but also systematic one induced by unclear physical quantities is required to confirm the reliability of calculated results. In this study, we applied a method based on analysis of variance to shielding calculations. We proposed random- and three-condition methods. The first one determines randomly the value of the unclear quantity, while the second one uses only three values: the default value, upper and lower limits. The systematic uncertainty can be estimated adequately by the random-condition method, though it needs the large computational cost. The three-condition method can provide almost the same estimate as the random-condition method when the effect of the variation is monotonic. We found criterion to confirm convergence of the systematic uncertainty as the number of trials increases.
Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa
JAEA-Data/Code 2016-018, 421 Pages, 2017/03
JAEA-Data-Code-2016-018.pdf:3.89MB
JAEA-Data-Code-2016-018-appendix(CD-ROM).zip:4.02MBJAEA-Data-Code-2016-018-hyperlink.zip:1.94MB
In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants, etc. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions.
Iwamoto, Yosuke; Taniguchi, Shingo*; Nakao, Noriaki*; Itoga, Toshio*; Nakamura, Takashi*; Nakane, Yoshihiro; Nakashima, Hiroshi; Satoh, Daiki; Yashima, Hiroshi*; Yamakawa, Hiroshi*; et al.
Nuclear Instruments and Methods in Physics Research A, 562(2), p.789 - 792, 2006/06
Times Cited Count：6 Percentile：43.74(Instruments & Instrumentation)Neutron energy spectra produced from thick targets play an important role in validation of calculation codes that are employed in the design of spallation neutron sources and the shielding design of accelerator facilities. However, appropriate experimental data were scarce in the forward direction for the incident energy higher than 100 MeV. In this study, neutron spectra at 0 degree from thick targets bombarded with 350 MeV protons were measured by the time-of-flight technique using an NE213. The targets used were graphite, Al, Fe and Pb and their thicknesses were chosen to be a little thicker than the stopping lengths. The experiment was carried out at the TOF course of the RCNP (Research Center of Nuclear Physics) ring cyclotron, Osaka University. The flight path length between center of the target and of an NE213 were 11.4 m for the measurement of low energy neutrons and 95 m for high energy neutrons. The experimental data are compared with the calculated results by using the Monte Carlo transport codes, such as MCNPX and PHITS codes.
Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki
JAERI 1348, 388 Pages, 2005/06
To realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector supercomputers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them.
Mori, Takamasa; Okumura, Keisuke; Nagaya, Yasunobu
Transactions of the American Nuclear Society, 84, p.45 - 46, 2001/06
no abstracts in English
Mori, Takamasa; Okumura, Keisuke; Nagaya, Yasunobu
Advanced Monte Carlo for Radiation Physics, Particle Transport Simulation and Applications, p.625 - 630, 2001/00
no abstracts in English
Petkov, P.*; Takeda, Toshikazu*; Mori, Takamasa
Annals of Nuclear Energy, 26(10), p.935 - 942, 1999/00
no abstracts in English
; Sasa, Toshinobu; Takada, Hiroshi; Takizuka, Takakazu
Proc. of 2nd Int. Conf. on Accelerator-Driven Transmutation Technologies and Applications, 1, p.668 - 674, 1996/00
no abstracts in English
Mori, Takamasa; Nakakawa, Masayuki
JAERI-Data/Code 94-007, 152 Pages, 1994/08
JAERI-Data-Code-94-007.pdf:3.92MB
no abstracts in English
*; Nakakawa, Masayuki; Mori, Takamasa
Comput. Assist. Mech. Eng. Sci., 1, p.177 - 189, 1994/00
no abstracts in English
Mori, Takamasa; Nakakawa, Masayuki; *
Journal of Nuclear Science and Technology, 29(4), p.325 - 336, 1992/04
no abstracts in English
; ;
JAERI-M 87-010, 28 Pages, 1987/02
no abstracts in English
Nakagawa, Masayuki;
JAERI-M 84-126, 74 Pages, 1984/07
no abstracts in English
; ; R.T.Santoro*; *; *
Nucl.Technol./Fusion, 2, p.272 - 285, 1982/00
no abstracts in English
; ; R.T.Santoro*; *; *
Transactions of the American Nuclear Society, 38, p.555 - 557, 1981/00
no abstracts in English
Nakagawa, Masayuki;
JAERI-M 8556, 40 Pages, 1979/11
no abstracts in English
;
Journal of Nuclear Science and Technology, 14(8), p.603 - 609, 1977/08
Times Cited Count：0no abstracts in English
; ; ; ; ; ; ; ;
JAERI-M 5557, 32 Pages, 1974/02
no abstracts in English
Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa
no journal, ,
A general-purpose Monte Carlo code MVP has been developed for continuous-energy neutron and photon transport calculations since the late 1980s at Japan Atomic Energy Agency. The MVP code is designed for nuclear reactor applications such as reactor core design/analysis, criticality safety and reactor shielding. The code has been widely in domestic use since the first release in 1994 and the second release in 2005. Modifications and enhancements have been made with advanced Monte Carlo methodology for reactor physics applications. Featured capabilities for version 3 are the perturbation calculation for the k-effective value, treatment of delayed neutrons, group constant generation, exact resonance elastic scattering model, reactor kinetics parameter calculation. These capabilities are integrated into the code and MVP version 3 is planned to be domestically released in near future.
