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Okumura, Keisuke; Kugo, Teruhiko; Kaneko, Kunio*; Tsuchihashi, Keiichiro*
JAEA-Data/Code 2007-004, 313 Pages, 2007/02
JAEA-Data-Code-2007-004.pdf:13.5MB
The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, SN transport codes ANISN(1D) and TWOTRAN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation.
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.
Kaneko, Kunio*
JNC TJ9400 2003-003, 150 Pages, 2003/03
JNC-TJ9400-2003-003.pdf:5.69MB
A 900 group cross section library based on the specification determined last year was produced for 27 nuclei of the fast reactor benchmark problem evaluated in nuclear data file JENDL-3.2. In addition, the new SLAROM code, which has been developed as an advanced detail analysis system, was revised so as to make cell calculations effectively with the above 900 group library. Furthermore, new functions were added to the SLAROM so that the SLAROM evaluates assembly parameters using effective cross sections derived by the SLAROM and produces any condensed effective cross section set for core perfomance analysis. With the 900 group cross section library and the revised SLAROM, three cell calculations for fast and medium neutron speed reactors having different neutron spectrum were performed, and the results were compared with those calculated by the continuous energy Monte Carlo code MVP. By the comparisons, it is concluded that the newly revised SLAROM and a 900 group cross section library give accuracy comparable to MVP for predicting core perfomances.
Sasa, Toshinobu; Tsujimoto, Kazufumi; Kaneko, Kunio*; Takano, Hideki
Journal of Nuclear Science and Technology, 39(Suppl.2), p.1183 - 1186, 2002/08
no abstracts in English
Okumura, Keisuke; Nakakawa, Masayuki; Kaneko, Kunio*; *
JAERI-Conf 2000-018, p.31 - 41, 2001/01
Burnup calculation codes based on the conventional deterministic approach often encounter difficult problems because of the constraints on the geometry description, limit of approximation on the effective resonance cross-sections, failing of the diffusion approximation due to extremely strong anisotropic or heterogenity. They are, for example, the prediction of burn characteristics of plutonium spot, core design of ultra-small reactors, analysis of the sample material in an irradiation capsule of the research rector. To deal with these problems any time, a burn-up calculation code (MVP-BURN) was developed by using a continuous energy Monte Carlo code MVP. MVP-BURN was validated by comparison with the results of deterministic codes in the international benchmark problems, and by comparison with the measured values of the spent fuel composition irradiated in a commercial reactor.
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
Kugo, Teruhiko; Kaneko, Kunio*
Mathematics and Computation, Reactor Physics and Environmental Analysis in Nuclear Applications, 2, p.2113 - 2122, 1999/09
no abstracts in English
Takano, Hideki; Akie, Hiroshi; Kaneko, Kunio*
Proc. of Int. Conf. on the Phys. of Nucl. Sci. and Technol., 1, p.58 - 65, 1998/00
no abstracts in English
Okumura, Keisuke; Nakakawa, Masayuki; Kaneko, Kunio*
Proc. of SARATOGA 1997, 1, p.495 - 508, 1997/00
no abstracts in English
Takano, Hideki*; Kaneko, Kunio*; Ishiguro, Yukio*
PNC TJ9500 89-001, 75 Pages, 1989/03
The fast reactor 70-group constant set JFS-3-J3T has been generated by using the JENDL-3T nuclear data. One-dimensional benchmark calculations and the analyses of the ZPPR-9 and FCA-VI-2 assemblies were performed with the JFS-3-J3T set. The results obtained are summarized as follows: (1)The k
's are underestimated by 0.6% for Pu-fuel cores and overestimated by 2% for U-fuel cores. (2)The central reaction rate ratio 
(Pu-239)/(U-235)is in a good agreement with the experimental value and 
(U-238)/(Pu-239) is overestimated. (3)Doppler and Na-void reactivities are in a good agreement with the measured data. (4)The radial reaction rate distributions are improved in the comparison of the results obtained with the JENDL-2 data. Furthermore, the benchmark test of JENDL-3T/Rev.1 which was revised for several important nuclides on the basis of the results described above has been performed. It was shown that JENDL-3T/Rev.1 predicted nuclear characteristics more satisfactory than JENDL-2, excepting the overestimate for the reaction rate ratios of (U-238)/(Pu-239) and (U-238)/(U-235).
