Probabilistic transient analysis of fuel choices for sodium fast reactors
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011.
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| Format: | Thesis |
| Language: | eng |
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Massachusetts Institute of Technology
2013
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| Online Access: | http://hdl.handle.net/1721.1/76578 |
| _version_ | 1826198162879021056 |
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| author | Denman, Matthew R |
| author2 | Neil E. Todreas. |
| author_facet | Neil E. Todreas. Denman, Matthew R |
| author_sort | Denman, Matthew R |
| collection | MIT |
| description | Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011. |
| first_indexed | 2024-09-23T11:00:11Z |
| format | Thesis |
| id | mit-1721.1/76578 |
| institution | Massachusetts Institute of Technology |
| language | eng |
| last_indexed | 2024-09-23T11:00:11Z |
| publishDate | 2013 |
| publisher | Massachusetts Institute of Technology |
| record_format | dspace |
| spelling | mit-1721.1/765782019-04-12T07:32:54Z Probabilistic transient analysis of fuel choices for sodium fast reactors Denman, Matthew R Neil E. Todreas. Massachusetts Institute of Technology. Dept. of Nuclear Science and Engineering. Massachusetts Institute of Technology. Dept. of Nuclear Science and Engineering. Nuclear Science and Engineering. Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011. Cataloged from PDF version of thesis. "June 2011." Includes bibliographical references (p. 180-184). This thesis presents the implications of using a risk-informed licensing framework to inform the design of Sodium Fast Reactors. NUREG-1860, more commonly known as the Technology Neutral Framework (TNF), is a risk-informed licensing process drafted by the Nuclear Regulatory Commission's (NRC) Office of Nuclear Regulatory Research. The TNF determines the acceptability of accident sequences by examining the 95th percentile estimate of both the frequency and quantity of radioactive material release and compares this value to predetermined limits on a Frequency-Consequence Curve. In order to apply this framework, two generic pool type sodium reactors, one using metal fuel and one using oxide fuel, were modeled in RELAP5-3D in order to determine the transient response of reactors to unprotected transient overpower and unprotected loss of flow events. Important transient characteristics, such as the reactivity coefficients, were treated as random variables which determine the success or failure of surviving the transient. In this context, success is defined as the cladding remaining intact and the avoidance of sodium boiling. In order to avoid running an excessive amount of simulations, the epistemic uncertainties around the random variables are sampled using importance sampling. For metallic fuel, the rate of fuel/cladding eutectic formation has typically been modeled as an Arrhenius process which depends only on the temperature of the fuel/cladding interface. Between the 1960s and the 1990s, numerous experiments have been conducted which indicate that the rate of fuel/cladding eutectic formation is more complex, depending upon fuel/cladding interfacial temperature, fuel constituents (uranium metal or uranium zirconium), cladding type (stainless steel 316, D9 or HT9), linear power, plutonium enrichment and burnup. This thesis improves the modeling accuracy of eutectic formation through the application of multivariable regression using a database of fuel/cladding eutectic experiments and determines that the remaining uncertainty governing the rate of eutectic formation should not significantly affect the frequency of cladding failure for tested cladding options. The general conclusion from this thesis is that when using NUREG-1860 to license metal or oxide fueled SFRs, it is steady state, not transient, cladding considerations which control optimal operating temperature, currently corresponding to an approximate core outlet temperature of 550°C. Metallic cores traditionally have been designed with core outlet temperatures of 510°C and increasing this temperature to 550°C may decrease the busbar cost by 19% when combined with the adoption of a Supercritical-CO₂ power conversion cycle, reduced containment requirements, and Printed Circuit Heat Exchangers. While both fuel types will be shown to meet the NUREG-1860 requirements, the frequency of radiation release for unprotected loss of flow and unprotected transient overpower events for metallic fuel has been shown to be orders of magnitude lower than for oxide fuel. by Matthew R. Denman. Ph.D. 2013-01-23T20:28:44Z 2013-01-23T20:28:44Z 2011 Thesis http://hdl.handle.net/1721.1/76578 823502450 eng M.I.T. theses are protected by copyright. They may be viewed from this source for any purpose, but reproduction or distribution in any format is prohibited without written permission. See provided URL for inquiries about permission. http://dspace.mit.edu/handle/1721.1/7582 xv, 244 p. application/pdf Massachusetts Institute of Technology |
| spellingShingle | Nuclear Science and Engineering. Denman, Matthew R Probabilistic transient analysis of fuel choices for sodium fast reactors |
| title | Probabilistic transient analysis of fuel choices for sodium fast reactors |
| title_full | Probabilistic transient analysis of fuel choices for sodium fast reactors |
| title_fullStr | Probabilistic transient analysis of fuel choices for sodium fast reactors |
| title_full_unstemmed | Probabilistic transient analysis of fuel choices for sodium fast reactors |
| title_short | Probabilistic transient analysis of fuel choices for sodium fast reactors |
| title_sort | probabilistic transient analysis of fuel choices for sodium fast reactors |
| topic | Nuclear Science and Engineering. |
| url | http://hdl.handle.net/1721.1/76578 |
| work_keys_str_mv | AT denmanmatthewr probabilistictransientanalysisoffuelchoicesforsodiumfastreactors |