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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.
Shinohara, Masanori; Ishitsuka, Etsuo; Shimazaki, Yosuke; Sawahata, Hiroaki
JAEA-Technology 2016-033, 65 Pages, 2017/01
JAEA-Technology-2016-033.pdf:11.14MB
To reduce the neutron exposure dose for workers during the replacement works of the startup neutron sources of the High Temperature Engineering Test Reactor, calculations of the exposure dose in case of temporary neutron shielding at the bottom of fuels handling machine were carried out by the PHITS code. As a result, it is clear that the dose equivalent rate due to neutron radiation can be reduced to about an order of magnitude by setting a temporary neutron shielding at the bottom of shielding cask for the fuel handling machine. In the actual replacement works, by setting temporary neutron shielding, it was achieved that the cumulative equivalent dose of the workers was reduced to 0.3 man mSv which is less than half of cumulative equivalent dose for the previous replacement works; 0.7 man mSv.
Okumura, Keisuke
JAEA-Data/Code 2015-015, 162 Pages, 2015/10
JAEA-Data-Code-2015-015.pdf:3.99MBJAEA-Data-Code-2015-015-appendix(CD-ROM).zip:3.38MB
MOSRA-SRAC is a lattice calculation module of the Modular code System for nuclear Reactor Analyses (MOSRA). This module performs the neutron transport calculation for various types of fuel elements including existing light water reactors, research reactors, etc. based on the collision probability method with a set of the 200-group cross-sections generated from the Japanese Evaluated Nuclear Data Library JENDL-4.0. It has also a function of the isotope generation and depletion calculation for up to 234 nuclides in each fuel material in the lattice. In these ways, MOSRA-SRAC prepares the burn-up dependent effective microscopic and macroscopic cross-section data to be used in core calculations.
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.
Kishimoto, Katsumi; Arigane, Kenji*
JAERI-Tech 2005-016, 83 Pages, 2005/03
JAERI-Tech-2005-016.pdf:10.52MB
Revaluation to activation activity of reactor evaluated at the notification of dismantling submitted in 1997 was carried out in JRR-2 where decommissioning was advanced now. In the revaluation, estimation accuracy on neutron streaming at various horizontal experimental tubes was improved by applying 3 dimensional model to neutron transport calculation that had been carried out by 2 dimensional model, and calculating with TORT. As the result, excessive overestimations on horizontal experimental tubes and biological shield that had greatly contributed to total activation activity in evaluation at the notification of dismantling was revised, sum of their activation activities in the revaluation decreased to 1/18(case after 1 year from the permanent shutdown of reactor) of evaluation at the notification of dismantling, and the structural materials that had large activation activity were changed. By the above, it was shown that introducing 3 dimensional model was effective in evaluation on activation activity of the research reactor that had a lot of various experimental tubes.
Sakamoto, Yukio
JAERI-Research 2002-025, 34 Pages, 2002/11
JAERI-Research-2002-025.pdf:1.91MB
no abstracts in English
Kugo, Teruhiko
Journal of Nuclear Science and Technology, 39(3), p.256 - 263, 2002/03
Times Cited Count：3 Percentile：23.2(Nuclear Science & Technology)Two vector computation algorithms; an odd-even sweep (OES) method and an independent sequential sweep (ISS) method, have been developed for the characteristics method to solve the neutron transport equation in a heterogeneous geometry. They realize long vector lengths without recursive operations for effective use of vector computers. Their efficiency has been investigated to a realistic fuel assembly calculation. For both methods, a vector computation is 15 times faster than a scalar computation. From a viewpoint of a comparison between the OES and ISS methods, the ISS method is superior to the OES method because the ISS method shows a faster convergence and saves a computer memory without reducing a computation speed.
Research Committee on Reactor Physics
JAERI-Review 2001-047, 180 Pages, 2002/02
JAERI-Review-2001-047.pdf:10.03MB
Under the Research Committee on Reactor Physics, the Working Party on Reactor Physics of Accelerator-Driven System (ADS-WP) was set in July 1999 to review and investigate special subjects related to reactor physics research for the Accelerator-Driven Subcritical System (ADS).The ADS-WP, at the first meeting, discussed a task guideline of its activity for two years and decided to concentrate upon three subjects: (1) neutron transport calculations in high energy range, (2) static and kinetic (safety-related) characteristics of subcritical system, and (3) system design including ADS concepts and elemental technology developments required.The activity of ADS-WP continued from July 1999 to March 2001. In this duration, the members of ADS-WP met together four times and discussed the above subjects. In addition, the ADS-WP conducted a questionnaire on requests and proposals for the plan of Transmutation Physics Experimental Facility in the High-Intensity Proton Accelerator Project, which is a joint project between JAERI and KEK (High Energy Accelerator Research Organization).This report summarizes the results obtained by the above ADS-WP activity. The report will be useful to overview those results and moreover to set up a new guideline of future research activity in this field.
Kugo, Teruhiko
JAERI-Research 2001-051, 39 Pages, 2001/11
JAERI-Research-2001-051.pdf:2.04MB
Fast computation of the characteristics method to solve the neutron transport equation in a heterogeneous geometry has been studied. Two vector computation algorithms; an odd-even sweep (OES) method and an independent sequential sweep (ISS) method have been developed and their efficiency to a typical fuel assembly calculation has been investigated. For both methods, a vector computation is 15 times faster than a scalar computation. From a viewpoint of comparison between the OES and ISS methods, the ISS method is superior to the OES method because the ISS method shows a faster convergence and saves a computer memory without reducing a computation speed. In the vector computation, a table-look-up method to reduce computation time of an exponential function saves only 20% of a whole computation time. Both the coarse mesh rebalance method and the Aitken acceleration method are effective as acceleration methods for the characteristics method, a combination of them saves 70-80% of outer iterations compared with a free iteration.
Sukegawa, Takenori; Hatakeyama, Mutsuo; Yanagihara, Satoshi
JAERI-Tech 2001-058, 81 Pages, 2001/09
JAERI-Tech-2001-058.pdf:5.98MB
In general, neutron transport and activation calculation codes are used for residual radioactive inventory estimation; however, it is essential to verify calculations by measurement results because of geometrical complexity of the reactor and so on. The comparison between measured and calculated radioactivity in the JPDR core components showed a relatively good agreement (factor of 2), and it was cleared that water content and weight ratio of steel bars to concrete materials significantly influenced the neutron flux distribution in the biological shield (factor of 2-10 error). The measured radioactivity inside of the reactor pressure vessel wall and at the inner part of the biological shield was compared in detail with the calculations to verify the methodology applied to calculations of radioisotope production. Then it was found that the radioactive inventory could be estimated accurately with combination of calculations and measurement of radioactivity in samples and dose rate distribution for planning of dismantling activities.
Endo, Akira; Yamaguchi, Yasuhiro; Sakamoto, Yukio; Yoshizawa, Michio; Tsuda, Shuichi
Radiation Protection Dosimetry, 93(3), p.207 - 214, 2001/00
Times Cited Count：6 Percentile：44.04(Environmental Sciences)no abstracts in English
Maekawa, Fujio; Konno, Chikara; Oyama, Yukio; Uno, Yoshitomo; Maekawa, Hiroshi; Ikeda, Yujiro
Fusion Engineering and Design, 42, p.275 - 280, 1998/00
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
Okumura, Keisuke; *;
JAERI-Data/Code 96-015, 445 Pages, 1996/03
JAERI-Data-Code-96-015.pdf:12.94MB
no abstracts in English
S.Zimin*; *; Takatsu, Hideyuki; Sato, Satoshi; Tsunematsu, Toshihide; Inoue, Takashi; Ohara, Yoshihiro
JAERI-Tech 94-015, 37 Pages, 1994/08
no abstracts in English
Fujimura, Toichiro
Computer Physics Communications, 82, p.111 - 119, 1994/00
Times Cited Count：1 Percentile：21.58(Computer Science, Interdisciplinary Applications)no abstracts in English
Oyama, Yukio; Maekawa, Hiroshi;
Nihon Genshiryoku Gakkai-Shi, 36(7), p.612 - 618, 1994/00
no abstracts in English
S.Zimin*; Takatsu, Hideyuki; Mori, Seiji*; *; Seki, Yasushi; Sato, Satoshi; Tada, Eisuke
JAERI-M 93-141, 75 Pages, 1993/07
no abstracts in English
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JAERI-M 87-010, 28 Pages, 1987/02
no abstracts in English
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JAERI-M 86-028, 107 Pages, 1986/03
no abstracts in English
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JAERI-M 85-086, 40 Pages, 1985/07
no abstracts in English
