Full-core analysis for FeCrAl enhanced accident tolerant fuel in boiling water reactors
Abstract
The impact of replacing Zircaloy with FeCrAl, a candidate enhanced accident-tolerant fuel cladding material, was evaluated for 10 × 10 boiling water reactor fuel bundles. Results from a series of full-core parametric studies estimated that replacing UO2/Zircaloy with UO2/FeCrAl would require an average enrichment increase of 0.6% 235U throughout the fuel lattice with the cladding and channel box thicknesses halved and fuel pellet diameter increased. Full-core results indicated that UO2/FeCrAl models with these geometric/enrichment specifications matched the base UO2/Zircaloy cycle length of 527 effective full power days. Optimization studies of the full-core design established loading and control blade patterns for both Zircaloy and FeCrAl models. A side study was conducted modeling a hybrid fuel bundle consisting of FeCrAl cladding and a SiC/Ni/Cr channel box. By halving the cladding thickness, the enrichment level required was less than that of the Zircaloy base case design after performing loading pattern optimization of the hybrid bundle core. Lastly, the thermomechanical performance of a Zircaloy-cladded fuel rod was compared to that of a FeCrAl system. Results from this analysis show that, if starting from the same fuel-cladding gap thickness, a FeCrAl-clad fuel rod operates with a greater average fuel centerline temperature, comparable axial elongation andmore »
- Authors:
-
[2];
[2];
[2];
- Univ. of Tennessee, Knoxville, TN (United States)
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Publication Date:
- Research Org.:
- Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
- Sponsoring Org.:
- USDOE
- OSTI Identifier:
- 1515695
- Alternate Identifier(s):
- OSTI ID: 2325462
- Grant/Contract Number:
- AC05-00OR22725
- Resource Type:
- Accepted Manuscript
- Journal Name:
- Annals of Nuclear Energy (Oxford)
- Additional Journal Information:
- Journal Name: Annals of Nuclear Energy (Oxford); Journal Volume: 132; Journal Issue: C; Journal ID: ISSN 0306-4549
- Publisher:
- Elsevier
- Country of Publication:
- United States
- Language:
- English
- Subject:
- 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; Accident tolerant fuel; BWR; FeCrAl; Alternate cladding; Full core
Citation Formats
George, Nathan M., Sweet, Ryan T., Powers, Jeffrey J., Worrall, Andrew, Terrani, Kurt A., Wirth, Brian D., and Maldonado, G. Ivan. Full-core analysis for FeCrAl enhanced accident tolerant fuel in boiling water reactors. United States: N. p., 2019.
Web. doi:10.1016/j.anucene.2019.04.025.
George, Nathan M., Sweet, Ryan T., Powers, Jeffrey J., Worrall, Andrew, Terrani, Kurt A., Wirth, Brian D., & Maldonado, G. Ivan. Full-core analysis for FeCrAl enhanced accident tolerant fuel in boiling water reactors. United States. https://doi.org/10.1016/j.anucene.2019.04.025
George, Nathan M., Sweet, Ryan T., Powers, Jeffrey J., Worrall, Andrew, Terrani, Kurt A., Wirth, Brian D., and Maldonado, G. Ivan. Mon .
"Full-core analysis for FeCrAl enhanced accident tolerant fuel in boiling water reactors". United States. https://doi.org/10.1016/j.anucene.2019.04.025. https://www.osti.gov/servlets/purl/1515695.
@article{osti_1515695,
title = {Full-core analysis for FeCrAl enhanced accident tolerant fuel in boiling water reactors},
author = {George, Nathan M. and Sweet, Ryan T. and Powers, Jeffrey J. and Worrall, Andrew and Terrani, Kurt A. and Wirth, Brian D. and Maldonado, G. Ivan},
abstractNote = {The impact of replacing Zircaloy with FeCrAl, a candidate enhanced accident-tolerant fuel cladding material, was evaluated for 10 × 10 boiling water reactor fuel bundles. Results from a series of full-core parametric studies estimated that replacing UO2/Zircaloy with UO2/FeCrAl would require an average enrichment increase of 0.6% 235U throughout the fuel lattice with the cladding and channel box thicknesses halved and fuel pellet diameter increased. Full-core results indicated that UO2/FeCrAl models with these geometric/enrichment specifications matched the base UO2/Zircaloy cycle length of 527 effective full power days. Optimization studies of the full-core design established loading and control blade patterns for both Zircaloy and FeCrAl models. A side study was conducted modeling a hybrid fuel bundle consisting of FeCrAl cladding and a SiC/Ni/Cr channel box. By halving the cladding thickness, the enrichment level required was less than that of the Zircaloy base case design after performing loading pattern optimization of the hybrid bundle core. Lastly, the thermomechanical performance of a Zircaloy-cladded fuel rod was compared to that of a FeCrAl system. Results from this analysis show that, if starting from the same fuel-cladding gap thickness, a FeCrAl-clad fuel rod operates with a greater average fuel centerline temperature, comparable axial elongation and radial displacement, and longer time to gap closure compared to a Zircaloy-clad fuel rod. Furthermore, this fuel performance analysis was primarily based on the commercial Kanthal APMT FeCrAl alloy but also used available data for the C35M FeCrAl alloy developed at Oak Ridge National Laboratory.},
doi = {10.1016/j.anucene.2019.04.025},
journal = {Annals of Nuclear Energy (Oxford)},
number = C,
volume = 132,
place = {United States},
year = {Mon May 13 00:00:00 EDT 2019},
month = {Mon May 13 00:00:00 EDT 2019}
}
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