Reactor Neutronics: Impact of Fissile Material
Abstract
Here, given a wide variety of reactor designs and fuel types, it can be difficult to identify the underlying cause of basic performance differences such as flux level and enrichment requirement. In this paper, using solely the definitions of the core multiplication factor and core power, simple relations have been derived allowing estimates of the flux ratio and fissile material concentration ratio for any reactor concept when 235U is replaced with 239Pu or vice-versa. These relations are functions of the neutron non-leakage probability, and one only needs to know number of neutrons emitted per fission, and the fission cross-section ratio between the 235U system and the 239Pu system. It is found that for a reactor concept having significant leakage, the achievable flux level when using 239Pu as fissile material can be up to 45% larger than when using 235U as fissile material, and the required fissile concentration of 239Pu is up to 48% lower than that of 235U to achieve criticality.
- Authors:
-
- Argonne National Lab. (ANL), Argonne, IL (United States)
- Publication Date:
- Research Org.:
- Argonne National Lab. (ANL), Argonne, IL (United States)
- Sponsoring Org.:
- USDOE Office of Nuclear Energy (NE)
- OSTI Identifier:
- 1393937
- Grant/Contract Number:
- AC02-06CH11357
- Resource Type:
- Accepted Manuscript
- Journal Name:
- Nuclear Science and Engineering
- Additional Journal Information:
- Journal Volume: 187; Journal Issue: 2; Journal ID: ISSN 0029-5639
- Publisher:
- American Nuclear Society - Taylor & Francis
- Country of Publication:
- United States
- Language:
- English
- Subject:
- 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; fissile material; neutron flux level; neutronics
Citation Formats
Heidet, F., and Hill, R. N. Reactor Neutronics: Impact of Fissile Material. United States: N. p., 2017.
Web. doi:10.1080/00295639.2017.1312933.
Heidet, F., & Hill, R. N. Reactor Neutronics: Impact of Fissile Material. United States. https://doi.org/10.1080/00295639.2017.1312933
Heidet, F., and Hill, R. N. Fri .
"Reactor Neutronics: Impact of Fissile Material". United States. https://doi.org/10.1080/00295639.2017.1312933. https://www.osti.gov/servlets/purl/1393937.
@article{osti_1393937,
title = {Reactor Neutronics: Impact of Fissile Material},
author = {Heidet, F. and Hill, R. N.},
abstractNote = {Here, given a wide variety of reactor designs and fuel types, it can be difficult to identify the underlying cause of basic performance differences such as flux level and enrichment requirement. In this paper, using solely the definitions of the core multiplication factor and core power, simple relations have been derived allowing estimates of the flux ratio and fissile material concentration ratio for any reactor concept when 235U is replaced with 239Pu or vice-versa. These relations are functions of the neutron non-leakage probability, and one only needs to know number of neutrons emitted per fission, and the fission cross-section ratio between the 235U system and the 239Pu system. It is found that for a reactor concept having significant leakage, the achievable flux level when using 239Pu as fissile material can be up to 45% larger than when using 235U as fissile material, and the required fissile concentration of 239Pu is up to 48% lower than that of 235U to achieve criticality.},
doi = {10.1080/00295639.2017.1312933},
journal = {Nuclear Science and Engineering},
number = 2,
volume = 187,
place = {United States},
year = {Fri Jun 09 00:00:00 EDT 2017},
month = {Fri Jun 09 00:00:00 EDT 2017}
}
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Works referenced in this record:
ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data
journal, December 2011
- Chadwick, M. B.; Herman, M.; Obložinský, P.
- Nuclear Data Sheets, Vol. 112, Issue 12
Core design studies for a 1000MWth Advanced Burner Reactor
journal, April 2009
- Kim, T. K.; Yang, W. S.; Grandy, C.
- Annals of Nuclear Energy, Vol. 36, Issue 3