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Jin, Tomoyuki*; Oki, Shigeo
JNC TN9410 2004-001, 79 Pages, 2004/03
ORLIBJ32 is a set of ORIGEN2 cross-section libraries for light water reactors and fast reactors based on Japanese evaluated nuclear data library JENDL-3.2. It has been opened since 1999. Following the latest revision of JENDL-3.2 to JENDL-3.3 in 2002, we have prepared new ORIGRN2 libraries for fast reactors. By using the same tool that made the fast reactor libraries in ORLIBJ32, the 73-group infinitely-diluted cross sections for 327 nuclides generated from JENDL-3.3 were collapsed into 1-group cross sections with an arbitrary weighting spectrum. For main nuclides, we used the 1-group shielded cross sections obtained by the fast reactor group constant set JFS-3-J3.3. As an isomeric ratio (g/(g+m)) for 241Am capture reaction, the value 0.85 was used instead of the conventional value of 0.80 in consequence of the latest research development of nuclear data. The new libraries were prepared for the following sodium-cooled fast reactors just like ORLIBJ32: JOYO (MK-I), MONJU, several kinds of a prototype reactor (600 MWe) parameterized by both fuel type (MOX, Metal, Nitride) and Pu isotopic composition, a commercial-size reactor (1300 MWe), and a Pu burning reactor. We performed burnup calculations with the new ORIGEN2 libraries in order to investigate the effect caused by the revision of the library.
Igarashi, Minoru; Tanaka, Masaaki; Kimura, Nobuyuki; Nakane, Shigeru*; Kawashima, Shigeyo*; Hayashi, Kenji; Tobita, Akira; Kamide, Hideki
JNC TN9400 2003-092, 100 Pages, 2003/11
A water experiment for thermal hydraulics in a mixing tee was performed to investigate thermal striping phenomena. Measurement of flow velocity using particle image velocimetry and temperature measurement were carried out. Normalized power spectrum density of temperature fluctuation had same profile, when the momentum ratio of the main and branch pips is the same. From the velocity measurement test, when the momentum ratio is the same, flow pattern at mixing region shows the alomost same tendency. Temperature transfer characteristics from fluid to structure can be estimated by a constant heat transfer coefficient in time.
Yamano, Hidemasa; Fujita, Satoshi; Tobita, Yoshiharu; Kondo, Satoru; Morita, Koji*; Sugaya, Masaaki*; Mizuno, Masahiro*; Hosono, Seigo*; Kondo, Teppei*
JNC TN9400 2003-070, 333 Pages, 2003/08
An advanced safety analysis computer code, SIMMER-III, has been developed at Japan Nuclear Cycle Development Institute (JNC) to more realistically investigate postulated core disruptive accidents in liquid-metal fast reactors. The two-dimensional framework of SIMMER-III fluid dynamics has been extended to three dimensions to a new code, SIMMER-IV, which is currently (in Version 2) coupled with the three-dimensional neutronics model. With the completion of the SIMMER-IV version, the applicability of the code is further enhanced and the many of the known limitations in SIMMER-III are eliminated. The sample calculations demonstrated the general validity of SIMMER-IV. This report describes SIMMER-IV Version 2.A, by documenting the models, numerical algorithms and code features, along with the program description and input and output information to aid the users.