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Journal Articles

Thermal-hydraulics technological strategy roadmap 2017; An Approach for continuous safety improvement of LWRs

Itoi, Tatsuya*; Iwaki, Chikako*; Onuki, Akira*; Kito, Kazuaki*; Nakamura, Hideo; Nishida, Akemi; Nishi, Yoshihisa*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 60(4), p.221 - 225, 2018/04

no abstracts in English

JAEA Reports

MVP/GMVP version 3; General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

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:4.02MB
JAEA-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.

Journal Articles

Thermal-hydraulics technological strategy roadmap that improves safety of LWRs

Arai, Kenji*; Umezawa, Shigemitsu*; Oikawa, Hirohide*; Onuki, Akira*; Nakamura, Hideo; Nishi, Yoshihisa*; Fujii, Tadashi*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 58(3), p.161 - 166, 2016/03

no abstracts in English

JAEA Reports

SWAT4.0; The Integrated burnup code system driving continuous energy Monte Carlo codes MVP, MCNP and deterministic calculation code SRAC

Kashima, Takao; Suyama, Kenya; Takada, Tomoyuki*

JAEA-Data/Code 2014-028, 152 Pages, 2015/03

JAEA-Data-Code-2014-028.pdf:13.39MB

There have been two versions of SWAT depending on details of its development history: the revised SWAT that uses the deterministic calculation code SRAC as a neutron transportation solver, and the SWAT3.1 that uses the continuous energy Monte Carlo code MVP or MCNP5 for the same purpose. It takes several hours, however, to execute one calculation by the continuous energy Monte Carlo code even on the super computer of the Japan Atomic Energy Agency. Moreover, two-dimensional burnup calculation is not practical using the revised SWAT because it has problems on production of effective cross section data and applying them to arbitrary fuel geometry when a calculation model has multiple burnup zones. Therefore, SWAT4.0 has been developed by adding, to SWAT3.1, a function to utilize the deterministic code SARC2006, which has shorter calculation time, as an outer module of neutron transportation solver for burnup calculation. SWAT4.0 has been enabled to execute two-dimensional burnup calculation by providing an input data template of SRAC2006 to SWAT4.0 input data, and updating atomic number densities of burnup zones in each burnup step. This report describes outline, input data instruction, and examples of calculations of SWAT4.0.

JAEA Reports

MVP/GMVP 2; General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki

JAERI 1348, 388 Pages, 2005/06

JAERI-1348.pdf:2.02MB

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.

JAEA Reports

Production of MVP neutron cross section libraries based on the latest evaluated nuclear data files

Mori, Takamasa; Nagaya, Yasunobu; Okumura, Keisuke; Kaneko, Kunio*

JAERI-Data/Code 2004-011, 119 Pages, 2004/07

JAERI-Data-Code-2004-011.pdf:5.93MB

The 2nd version of code system, LICEM-2, has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system can process nuclear data in the latest ENDF-6 format and produce cross section libraries for MVP's capability of transport calculation at arbitrary temperature. By using the present system, MVP neutron cross section libraries have been prepared from the latest evaluations of JENDL, ENDF/B and JEFF data bases. This report describes the specification of MVP neutron cross section library, the details of each code in the code system, how to use them and MVP neutron cross section libraries produced with the code system.

Journal Articles

Validation of integrated burnup code system SWAT2 by the analyses of isotopic composition of spent nuclear fuel

Suyama, Kenya; Mochizuki, Hiroki*; Okuno, Hiroshi; Miyoshi, Yoshinori

Proceedings of International Conference on Physics of Fuel Cycles and Advanced Nuclear Systems; Global Developments (PHYSOR 2004) (CD-ROM), 10 Pages, 2004/04

This paper provides validation results of SWAT2, the revised version of SWAT, which is a code system combining point burnup code ORIGEN2 and continuous energy Monte Carlo code MVP, by the analysis of post irradiation examinations (PIEs). Some isotopes show differences of calculation results between SWAT and SWAT2. However, generally, the differences are smaller than the error of PIE analysis that was reported in previous SWAT validation activity, and improved results are obtained for several important fission product nuclides. This study also includes comparison between an assembly and a single pin cell geometry models.

Journal Articles

Validation of a continuous-energy Monte Carlo burn-up code MVP-BURN and its application to analysis of post irradiation experiment

Okumura, Keisuke; Mori, Takamasa; Nakakawa, Masayuki; Kaneko, Kunio*

Journal of Nuclear Science and Technology, 37(2), p.128 - 138, 2000/02

no abstracts in English

Journal Articles

Monte Carlo analysis of HTTR with the MVP statistical geometry model

Mori, Takamasa; Okumura, Keisuke; Nagaya, Yasunobu; Ando, Hiroei

Transactions of the American Nuclear Society, 83, p.283 - 284, 2000/00

no abstracts in English

Journal Articles

Application of continuous energy Monte Carlo code MVP to burn-up and whloe core calculations using cross sections at arbitrary temperatures

Mori, Takamasa; Okumura, Keisuke; Nagaya, Yasunobu; Nakakawa, Masayuki

Mathematics and Computation, Reactor Physics and Environmental Analysis in Nuclear Applications, 2, p.987 - 996, 1999/09

no abstracts in English

Journal Articles

Development of burn-up calculation code system MVP-BURN based on continuous energy Monte Carlo method and its validation

Okumura, Keisuke; Nakakawa, Masayuki; Kaneko, Kunio*

Proc. of SARATOGA 1997, 1, p.495 - 508, 1997/00

no abstracts in English

JAEA Reports

Neutron cross section library production code system for continuous energy Monte Carlo code MVP; LICEM

Mori, Takamasa; Nakakawa, Masayuki; *

JAERI-Data/Code 96-018, 121 Pages, 1996/05

JAERI-Data-Code-96-018.pdf:3.38MB

no abstracts in English

Journal Articles

Continuous energy Monte Carlo calculations of randomly distributed spherical fuels in high-temperature gas-cooled reactors based on a statistical geometry model

Murata, Isao; Mori, Takamasa; Nakakawa, Masayuki

Nuclear Science and Engineering, 123, p.96 - 109, 1996/00

 Times Cited Count:34 Percentile:92.49(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Production and verification of the MCNP cross section library FSXLIB-J3R2 based on JENDL-3.2

Kosako, Kazuaki*; Yamano, Naoki*; Maekawa, Fujio; Oyama, Yukio

Proc., 1996 Topical Meeting on Radiation Protection and Shielding, 1, p.1088 - 1095, 1996/00

no abstracts in English

JAEA Reports

FSXLIB-J3R2: A continuous energy cross section library for MCNP based on JENDL-3.2

Kosako, Kazuaki*; Maekawa, Fujio; Oyama, Yukio; Uno, Yoshitomo; Maekawa, Hiroshi

JAERI-Data/Code 94-020, 42 Pages, 1994/12

JAERI-Data-Code-94-020.pdf:1.18MB

no abstracts in English

JAEA Reports

JAEA Reports

Journal Articles

Benchmark calculation for deep penetration problem of 14MeV neutrons in iron

Mori, Takamasa; Nakakawa, Masayuki

Journal of Nuclear Science and Technology, 29(11), p.1061 - 1073, 1992/11

no abstracts in English

Journal Articles

Vectorization of continuous energy Monte Carlo method for neutron transport calculation

Mori, Takamasa; Nakakawa, Masayuki; *

Journal of Nuclear Science and Technology, 29(4), p.325 - 336, 1992/04

no abstracts in English

22 (Records 1-20 displayed on this page)