Neutronic and thermal analysis of composite fuel for potential deployment in fast reactors
Description
Highlights: • Neutronic and heat transfer performance of composite fuels on the macro-scale. • Methodology to guide flexible fuel design using high fidelity simulation tools. • Viability of composite fuels for ultra-high burnup fast reactor deployment. - Abstract: Composite fuels are promising candidates for high-burnup fast reactors because of their accommodation of swelling, limited fuel-cladding interactions and flexibility in design. While a proof-of-concept fuel consisting of granules of U-alloys and PuO2 dispersed within a porous zirconium matrix was successfully manufactured and irradiated, its neutronic and thermal performance remains to be optimized as compared to currently utilized fuels. MCNP6, COMSOL and a sphere packing algorithm were employed to perform the analysis. We found that both the theoretical maximum burnup reached and the temperature profiles are comparable to that of the currently considered alternative fuel. The results are promising and do not indicate any substantial limitation to the deployment of composite fuel. The fuel type merits further research, including full-core simulations. The methodology followed herein also provides a basis for screening different material compositions and guiding materials selection in composite fuels.
Availability note (English)
Available from http://dx.doi.org/10.1016/j.nucengdes.2016.04.010Additional details
Identifiers
- DOI
- 10.1016/j.nucengdes.2016.04.010;
- PII
- S0029-5493(16)30028-0;
Publishing Information
- Journal Title
- Nuclear Engineering and Design
- Journal Volume
- 303
- Journal Page Range
- p. 50-57
- ISSN
- 0029-5493
- CODEN
- NEDEAU
INIS
- Country of Publication
- Netherlands
- Country of Input or Organization
- International Atomic Energy Agency (IAEA)
- INIS RN
- 48005742
- Subject category
- S42: ENGINEERING; S11: NUCLEAR FUEL CYCLE AND FUEL MATERIALS;
- Descriptors DEI
- ALGORITHMS; ALTERNATIVE FUELS; BURNUP; C CODES; CLADDING; COMPARATIVE EVALUATIONS; FAST REACTORS; FLEXIBILITY; FUEL-CLADDING INTERACTIONS; HEAT TRANSFER; IRRADIATION; M CODES; PLUTONIUM OXIDES; POROUS MATERIALS; SIMULATION; SWELLING; THERMAL ANALYSIS; URANIUM ALLOYS; ZIRCONIUM
- Descriptors DEC
- ACTINIDE ALLOYS; ACTINIDE COMPOUNDS; ALLOYS; CHALCOGENIDES; COMPUTER CODES; DEFORMATION; DEPOSITION; ELEMENTS; ENERGY TRANSFER; EPITHERMAL REACTORS; EVALUATION; FUELS; MATERIALS; MATHEMATICAL LOGIC; MECHANICAL PROPERTIES; METALS; OXIDES; OXYGEN COMPOUNDS; PLUTONIUM COMPOUNDS; REACTORS; SURFACE COATING; TENSILE PROPERTIES; TRANSITION ELEMENTS; TRANSURANIUM COMPOUNDS
Optional Information
- Copyright
- Copyright (c) 2016 Elsevier Science B.V., Amsterdam, The Netherlands, All rights reserved.