Verification of Nuclear Calculation Methodology and Preliminary Uncertainty Quantification in a Sodium-cooled Fast Reactor - 14126
Creators
- 1. Mitsubishi FBR systems, Inc., 34-17, Jingumae 2-chome, Shibuya-ku, Tokyo, 150-0001 (Japan)
- 2. Mitsubishi Heavy Industries, Ltd., 1-1, Wadamisaki-cho 1-chome, Hyogo-ku, Kobe, 652-8585 (Japan)
- 3. Japan Atomic Energy Agency, 4002, Narita, Oarai, Higashi-ibaraki-gun, Ibaraki, 311-1393 (Japan)
Description
This paper treats the verification of nuclear calculation methodology of control rod reactivity and the uncertainty quantification of computational model in order to design Japan Sodium-cooled Fast Reactor, a next demonstration fast breeder reactor. Verification and validation of design methodology is required and various kinds of uncertainty in nuclear characteristics should be comprehensively assessed. This study starts in pursuit of them in the context of preliminary conceptual design. First, this work compares the calculation results of a detailed deterministic calculation method with that of Monte Carlo one in order to verify it. The deterministic method uses the following codes: a three-dimensional triangle diffusion code, TRISTAN; a three-dimensional triangular prismatic discrete ordinates transport calculation code, ENSEMBLE-TRIZ; an ultrafine group cell calculation code based on the method of collision probability, SLAROM-UF; and a two-dimensional lattice cell calculation code based on the method of characteristics, GALAXY-H. The control rod reactivity is calculated in the 750 MWe sodium-cooled mixed-oxide-fuelled fast breeder reactor; the conditions of reactor core are at the beginning of initial core, and at the beginning and the end of equilibrium core. Comparing the deterministic calculation methodology with the simulations by a continuous energy Monte Carlo code, MVP, the discrepancies are unveiled to be 1.8% to 2.0% (primary control rod), and 1.6% to 2.4% (backup control rod). Second, the uncertainties associated with the computational model are preliminarily estimated to be 4.0% (primary control rod), and 4.2% (backup control rod), based on the correction values. Consequently, it is naturally concluded that the nuclear calculation methodology can prospect the control rod reactivity accurately within a few percentages, representing the mathematical model precisely. (authors)
Additional details
Publishing Information
- Publisher
- American Nuclear Society - ANS
- Imprint Place
- La Grange Park, IL (United States)
- ISBN
- 978-0-89448-776-7
- Imprint Pagination
- 9 p.
Conference
- Title
- International Congress on Advances in Nuclear Power Plants
- Acronym
- ICAPP 2014
- Dates
- 6-9 Apr 2014
- Place
- Charlotte, NC (United States)
INIS
- Country of Publication
- United States
- Country of Input or Organization
- France
- INIS RN
- 54022514
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S11: NUCLEAR FUEL CYCLE AND FUEL MATERIALS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- COMPUTERIZED SIMULATION; CONTROL ELEMENTS; DESIGN; DISCRETE ORDINATE METHOD; FBR TYPE REACTORS; MATHEMATICAL MODELS; MIXED OXIDE FUELS; MONTE CARLO METHOD; REACTIVITY; REACTOR CORES; SODIUM; SODIUM COOLED REACTORS; THREE-DIMENSIONAL LATTICES; TWO-DIMENSIONAL SYSTEMS
- Descriptors DEC
- ALKALI METALS; BREEDER REACTORS; CALCULATION METHODS; CRYSTAL LATTICES; CRYSTAL STRUCTURE; ELEMENTS; ENERGY SOURCES; EPITHERMAL REACTORS; FAST REACTORS; FUELS; LIQUID METAL COOLED REACTORS; MATERIALS; METALS; NUCLEAR FUELS; REACTOR COMPONENTS; REACTOR MATERIALS; REACTORS; SIMULATION; SOLID FUELS
Optional Information
- Notes
- 12 refs.; Available on CD-ROM from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US)