Verification of AEGIS/SCOPE2, a next-generation in-core fuel management system

Conference · Sat Jul 01 00:00:00 EDT 2006

OSTI ID:22039719

Tatsumi, M ^[1]; Sugimura, N ^[2]; Yamamoto, A ^[3]

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1. Nuclear Fuel Industries, Ltd., 1-950, Asashiro-Nishi, Kumatori-cho, Sennan-gun, Osaka 590-0481 (Japan)
2. Nuclear Engineering, Ltd., 1-3-7 Tosabori, Nishi-ku, Osaka 550-0001 (Japan)
3. Nagoya Univ., Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan)

AEGIS/SCOPE2 is a next-generation code system for in-core fuel management of PWRs; AEGIS is a 2-D lattice code which treats heterogeneous geometry based on the MOC, while SCOPE2 is a highly efficient parallel code which performs multi-group nodal-transport calculations in 3-D pin-by-pin geometry. Cross sections for SCOPE2 calculations are provided by AEGIS. In this paper, a preliminary result of numerical performance by the AEGIS/SCOPE2 system is presented. In assembly calculations, prediction results by AEGIS were compared with reference results by MVP, a continuous-energy Monte-Carlo code, for k{sub {infinity}} and fission rate distributions within an assembly. Good agreement between the codes was observed. A preliminary result of burnup calculation is also presented with comparisons of k{sub {infinity}} between AEGIS and MVP-burn, a burnup code coupled with MVP. AEGIS predicted k{sub {infinity}} within {+-}0.2 %{Delta}k/k throughout burnup up to 60 GWd/t compared to the reference. An initial core of a commercial PWR at HZP was analyzed with AEGIS/SCOPE2 using nuclear data libraries including ENDF-B/VI rev. 8, B/VTI beta 0 and JENDL-3.3. In this preliminary study, the criticality was a little underestimated, however assembly-wise power distribution was predicted in good accuracy. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22039719
• Resource Relation: Conference: PHYSOR-2006: American Nuclear Society's Topical Meeting on Reactor Physics - Advances in Nuclear Analysis and Simulation, Vancouver, BC (Canada), 10-14 Sep 2006; Other Information: Country of input: France; 11 refs.
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
97 MATHEMATICAL METHODS AND COMPUTING
A CODES
ACCURACY
BURNUP
COMPARATIVE EVALUATIONS
CRITICALITY
CROSS SECTIONS
FUEL MANAGEMENT
MONTE CARLO METHOD
MULTIGROUP THEORY
MULTIPLICATION FACTORS
NODAL EXPANSION METHOD
NUCLEAR DATA COLLECTIONS
POWER DISTRIBUTION
PWR TYPE REACTORS
REACTOR CORES
REACTOR LATTICE PARAMETERS
S CODES
THREE-DIMENSIONAL CALCULATIONS
TWO-DIMENSIONAL CALCULATIONS
VERIFICATION