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Okuno, Hiroshi; Takada, Tomoyuki; Yoshiyama, Hiroshi; Miyoshi, Yoshinori
JAEA-Data/Code 2005-001, 117 Pages, 2005/11
JAEA-Data-Code-2005-001.pdf:9.37MB
Criticality calculation codes/code systems MCNP, MVP, SCALE and JACS, which are currently typically used in Japan for nuclear criticality safety evaluation, were benchmarked for so called dissolver-typed systems, i.e., fuel rod arrays immersed in fuel solution. The benchmark analyses were made for the evaluated critical experiments published in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook: one evaluation representing five critical configurations from heterogeneous core of low-enriched uranium dioxides at the Japan Atomic Energy Research Institute and two evaluations representing 16 critical configurations from heterogeneous core of mixed uranium and plutonium dioxides (MOXs) at the Battelle Pacific Northwest Laboratories of the U.S.A.. The results of the analyses showed that the minimum values of the neutron multiplication factor obtained with MCNP, MVP, SCALE and JACS were 0.993, 0.990, 0.993, 0.972, respectively, which values are from 2% to 4% larger than the maximum permissible multiplication factor of 0.95.
Shiraishi, Kunio; Sukegawa, Takenori; Ishigami, Tsutomu
JAEA-Data/Code 2005-002, 162 Pages, 2006/01
JAEA-Data-Code-2005-002.pdf:27.87MB
In order to efficiently develop a decommissioning plan of a nuclear facility, it is useful to develop a database containing information on decommissioning technology, cost and risk analysis results, and decommissioning projects such as the JPDR decommissioning project by collecting the information systematically. A decommissioning database has been developed by collecting decommissioning related information and analyzing it. The database provides information on not only data of decommissioning technology and decommissioning projects but also laws and safety standards for decommissioning in each country and international organizations. The database is released in a Homepage on Web and is available for use via intranet with functions of retrieval, display and printing.
Suzuki, Motoe; Saito, Hiroaki*
JAEA-Data/Code 2005-003, 415 Pages, 2006/02
JAEA-Data-Code-2005-003.pdf:17.01MB
A light water reactor fuel analysis code FEMAXI-6 is an advanced version which has been produced by integrating the former version FEMAXI-V with numerous functional improvements and extensions. In particular, the FEMAXI-6 code has attained a complete coupled solution of thermal analysis and mechanical analysis, enabling an accurate prediction of pellet-clad gap size and PCMI in high burnup fuel rods. Also, such new models have been implemented as pellet-clad bonding and fission gas bubble swelling, and linkage function with detailed burning analysis code has been enhanced. Furthermore, a number of new materials properties and parameters have been introduced. With these advancements, the FEMAXI-6 code has been upgraded to a versatile analytical tool for high burnup fuel behavior not only in the normal operation but also in anticipated transient conditions. This report describes in detail the design, basic theory and structure, models and numerical method, improvements and extensions, and method of model modification. In order to facilitate effective and wide-ranging application of the code, formats and methods of input/output of the code are also described, and a sample output in an actual form is included.