3-D SYS-TH : an OECD/NEA activity on multi-dimensional capabilities of thermalhydraulic system
C. Herera, D. Bestionb, P. Fillionb, R. Preab, V.Parrinelloc, A. Bousbia Salahd, K. Kime, J. J. Jeongf
aIRSN Institut de Radioprotection et de Sureté Nucléaire BP 17 92262 Fontenay-aux-Roses Cedex France
bCEA, STMF, Université Paris-Saclay, 91191 Gif sur Yvette Cedex, France
cNINE N uclear and INdustrial Engineering, Borgo Giannotti, Lucca, Italy
dBelV, Belgium
eKAERI, South Korea
fPusan National University, South Korea
ICAPP 2019 - International Congress on Advances in Nuclear Power Plants
Juan-les-pins, France, 12 - 15 May 2019
Abstract — The evaluation model and computational capabilities required for engineering design and safety analyses of nuclear installations have shown significant progress compared to the first tools established in the 60s. Regarding thermalhydraulics, first generation system codes were based on simple one-dimensional models associated with conservatisms intended to cover lack of knowledge, simplifications and limited computational capabilities and limited experimental support available at that time.
The second generation, mainly one-dimensional with some limited multi-dimensional capabilities, implemented the more advanced two-fluid sixequation model, and adopted the best–estimate approach, benefited from an extensive experimental program, and showed large improvement compared to the first generation tools.
However, the current tools still have limitations that both industries and regulatory bodies would like to address. Next generation codes are being developed to achieve an improved thermal-hydraulic analysis capacity. Within the Working Group on Analysis and Management of Accidents (WGAMA) of the OECD/NEA, an activity has been initiated in 2016 aiming at establishing a state of the art of current 3D capabilities in thermal hydraulic system codes which covers all aspects and limitations, from the equations and simplifications considered, time and space averaging hypotheses with unavoidable use of relatively coarse meshes, closure models up to available or needed experimental support. The main findings of this activity are presented in this paper.
BENCHMARK OF FUEL PERFORMANCE CODES FOR FeCrAl CLADDING BEHAVIOR ANALYSIS
G. Pastore1 , K.A. Gamble1 , M. Cherubini2 , C. Giovedi3 , A. Marino4 , A. Yamaji5 , Y. Kaji6 , P. Van Uffelen7 ,M.S. Veshchunov8
1 Idaho National Laboratory, Idaho Falls, United States
2 Nuclear and Industrial Engineering, Lucca, Italy
3 University of São Paulo, São Paulo, Brazil
4 National Atomic Energy Commission, Bariloche, Argentina
5 Waseda University, Tokyo, Japan
6 Japan Atomic Energy Agency, Ibaraki, Japan
7 European Commission, Joint Research Centre, Directorate for Nuclear Safety and Security, Karlsruhe, Germany
8 International Atomic Energy Agency, Vienna, Austria
Top Fuel 2019, Seattle, WA, September 22-27, 2019
Abstract — Oxidation-resistant iron-chromium-aluminum (FeCrAl) steels have been proposed for application as cladding materials in light water reactor fuel rods with improved accident tolerance. Within the Coordinated Research Project ACTOF of the International Atomic Energy Agency
(IAEA), a fuel performance modeling benchmark for
FeCrAl cladding behavior was conducted. During this effort, calculations were performed with various fuel performance codes for a set of fuel rod problems with FeCrAl steel as cladding material, and results were compared to each other. Comparisons to Zircaloy-4 cladding behavior under the same conditions were also included. Considered cases covered both normal operating and loss of coolant conditions. Participating organizations were INL (BISON code), NINE (TRANSURANUS), JAEA in cooperation with Waseda University (FEMAXI-7), University of São Paulo (FRAPCON) and CNEA (BACO). Results of the benchmark are presented and discussed in this contribution.