Application of the JENDL-4.0 nuclear data set for uncertainty analysis of the prototype FBR Monju

Conference · Sun Jul 01 00:00:00 EDT 2012

OSTI ID:22107780

Tamagno, P. ^[1]; Takeda, T. ^[2]; Konomura, M. ^[3]

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1. Institut National des Sciences et Techniques Nucleaires, INSTN - Point Courrier No 35, Centre CEA de Saclay, F-91191 Gif-sur-Yvette Cedex (France)
2. Research Inst. of Nuclear Engineering, Univ. of Fukui, Bunkyo 3-9-1, T910-8507 Fukui-ken, Fukui-shi (Japan)
3. Japan Atomic Energy Agency, FBR Plant Engineering Center, Shiraki 1, 919-1279 Fukui-ken, Tsuruga-shi (Japan)

This paper deals with uncertainty analysis of the Monju reactor using JENDL-4.0 and the ERANOS code 1. In 2010 the Japan Atomic Energy Agency - JAEA - released the JENDL-4.0 nuclear data set. This new evaluation contains improved values of cross-sections and emphasizes accurate covariance matrices. Also in 2010, JAEA restarted the sodium-cooled fast reactor prototype Monju after about 15 years of shutdown. The long shutdown time resulted in a build-up of {sup 241}Am by natural decay from the initially loaded Pu. As well as improved covariance matrices, JENDL-4.0 is announced to contain improved data for minor actinides 2. The choice of Monju reactor as an application of the new evaluation seems then even more relevant. The uncertainty analysis requires the determination of sensitivity coefficients. The well-established ERANOS code was chosen because of its integrated modules that allow users to perform sensitivity and uncertainty analysis. A JENDL-4.0 cross-sections library is not available for ERANOS. Therefor a cross-sections library had to be made from the original ENDF files for the ECCO cell code (part of ERANOS). For confirmation of the newly made library, calculations of a benchmark core were performed. These calculations used the MZA and MZB benchmarks and showed consistent results with other libraries. Calculations for the Monju reactor were performed using hexagonal 3D geometry and PN transport theory. However, the ERANOS sensitivity modules cannot use the resulting fluxes, as these modules require finite differences based fluxes, obtained from RZ SN-transport or 3D diffusion calculations. The corresponding geometrical models have been made and the results verified with Monju restart experimental data 4. Uncertainty analysis was performed using the RZ model. JENDL-4.0 uncertainty analysis showed a significant reduction of the uncertainty related to the fission cross-section of Pu along with an increase of the uncertainty related to the capture cross-section of {sup 238}U compared with the previous JENDL-3.3 version. Covariance data recently added in JENDL-4.0 for {sup 241}Am appears to have a non-negligible contribution. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22107780
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
97 MATHEMATICS AND COMPUTING
AMERICIUM 241
BENCHMARKS
COMPUTER CODES
COMPUTER GRAPHICS
CROSS SECTIONS
DATA COVARIANCES
JAEA
MONJU REACTOR
NUCLEAR DATA COLLECTIONS
PLUTONIUM
REACTOR SHUTDOWN
SENSITIVITY
STATISTICAL MECHANICS
THREE-DIMENSIONAL CALCULATIONS
TRANSPORT THEORY
URANIUM 238