Preliminary neutronic study of the S-PRISM fast reactor
DOI:
https://doi.org/10.15392/bjrs.v8i3A.1580Keywords:
S-PRISM, fast reactor, neutronicAbstract
Waste management is a challenge currently faced by nuclear energy worldwide. Most nuclear power plants operate in a once-through cycle and all spent nuclear fuel is stored in long term. Fast reactors can recycle 96% of the material in spent nuclear fuel of thermal reactors, reducing drastically the time required to store the waste. The S-PRISM reactor is a modular, pool-type sodium-cooled fast reactor that can operate with a variety of fuel types. In this study, the advantages presented by fast reactors were briefly discussed and a preliminary analysis of the fuel pin cells was made. It was found that the isothermal coefficient of reactivity was lower than a regular PWR's, with neutron leakage playing a major role instead of coolant or doppler expansion. The k∞ showed decrease for the driver pin and increase for the blanket pin during burnup; opposite behavior was observed for the conversion ratio. No results for pin analysis were found in literature for comparison. As the simulations were carried out separately instead of in a system, the results are not representative of the whole core behavior but must be seen as a path to understand the different types of pin influence in the system.
- Views: 178
- PDF Downloads: 195
Downloads
References
IAEA. Management of Waste from the Use of Radioactive Material in Medicine, Industry, Agriculture, Research and Education. Safety Guide WSG2, International Atomic Energy Agency (2005).
Triplett B. S., Loewen E. P., Dooies B. J. Dooies. PRISM: a competitive small modular sodium-cooled reactor, Nuclear Technology, 178, pp. 186-200 (2012).
Hannum W. H., Wade D. C., McFarlane H. F., Hill, R. N. Non-proliferation and safeguards aspects of the IFR, Progress in Nuclear Energy, 31, pp. 203-217, (1997).
Dubberley A., Wu T., Kubo S. S-PRISM high burnup metal-fuel core design for session 3: Future deployment programs and issues. Proceeding of International Congress on Advances in Nuclear Power Plants - ICAPP, Cordoba, Spain, May 4-7 (2003).
Lamarsh J. R., Baratta A. J., Introduction to Nuclear Engineering, Prentice Hall (2001).
Sumner T., Ghiaasiaan S.M. Effects of fuel type on the safety characteristics of a sodium-cooled fast reactor: Part I: Background, modeling tools and pre-transient calculations, Annals of Nuclear Energy, 38, pp. 1559-1568, (2011).
LANL. A General Monte Calo N-Particle (MCNP) Transport Code, http://www.mcnp.lanl.gov (2019).
Brown D. A., Chadwick M. B., Capote R. et al. ENDF/B-VIII. 0: the 8th major release of the nuclear reaction data library with CIELO-project cross sections, new standards and thermal scattering data", Nuclear Data Sheets, 148, pp.1-142 (2018).
R. Macfarlane, D. W. Muir, R. M. Boicurt, A. C. Kahler III and J. L. Conlin, The NJOY Nuclear Data Processing System, Version 2016 (No. LA-UR-17-20093)". Los Alamos National Lab. (LANL), Los Alamos, NM (United States).
Published
How to Cite
Issue
Section
License
Copyright (c) 2021 Brazilian Journal of Radiation Sciences
This work is licensed under a Creative Commons Attribution 4.0 International License.
Licensing: The BJRS articles are licensed under a Creative Commons Attribution 4.0 International License, which permits use, sharing, adaptation, distribution and reproduction in any medium or format, as long as you give appropriate credit to the original author(s) and the source, provide a link to the Creative Commons license, and indicate if changes were made. The images or other third party material in this article are included in the article’s Creative Commons license, unless indicated otherwise in a credit line to the material. If material is not included in the article’s Creative Commons license and your intended use is not permitted by statutory regulation or exceeds the permitted use, you will need to obtain permission directly from the copyright holder. To view a copy of this license, visit http://creativecommons.org/licenses/by/4.0/
