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Title:Fuel processing simulation tool for liquid-fueled nuclear reactors
Author(s):Rykhlevskii, Andrei
Director of Research:Huff, Kathryn D
Doctoral Committee Chair(s):Huff, Kathryn D
Doctoral Committee Member(s):Kozlowski, Tomasz; Stubbins, James F; Olson, Luke N
Department / Program:Nuclear, Plasma, & Rad Engr
Discipline:Nuclear, Plasma, Radiolgc Engr
Degree Granting Institution:University of Illinois at Urbana-Champaign
Degree:Ph.D.
Genre:Dissertation
Subject(s):depletion
burnup
MSR
fuel
reprocessing
refueling
Monte Carlo
Abstract:Nuclear reactors with liquid fuel offer multiple advantages over their solid-fueled siblings: improved inherent safety, fuel utilization, thermal efficiency, online reprocessing, and potential for nuclear fuel cycle closure. To advance this promising reactor design, researchers need a simulation tool for fuel depletion calculations while taking into account online reprocessing and refueling. This work presents a flexible, open-source tool, SaltProc, for simulating the fuel depletion in a generic nuclear reactor with liquid, circulating fuel. SaltProc allows the user to define realistically constrained extraction efficiency of fission products based on physical models of fuel processing components appearing in various MSR (Molten Salt Reactor) systems. Developed using a Python Object-Oriented Programming paradigm, SaltProc can model a complex, multi-zone, multi-fluid MSR operation and is sufficiently general to represent myriad reactor systems. Moreover, SaltProc can maintain reactor criticality by adjusting the geometry of the core. Finally, the tool can analyze power variations in the context of depletion. This thesis also demonstrates and validates SaltProc for two prospective reactor designs: the Molten Salt Breeder Reactor (MSBR) and the Transatomic Power (TAP) MSR. A 60-year full-power MSBR depletion calculation with ideal fission product extraction (e.g., 100% of target poison removed) has been validated against Betzler et al. simulation results obtained with ChemTRITON at ORNL. The average 232Th feed rate obtained is the current work is 2.40 kg/d, which is consistent with ORNL results (2.45 kg/d). This simulation showed that the online fission product extraction and online refueling with 232Th allowed the MSBR to operate at full power for 60 years due to exceptionally low parasitic neutron absorption. This work shows fuel depletion simulations with SaltProc for the TAP MSR to demonstrate the tool capability to model liquid-fueled reactors with movable/adjustable moderator. This dissertation also validated depletion calculations for a realistic multi-component model of the fuel salt reprocessing system with assumed ideal extraction efficiency against full-core TAP depletion analysis by Betzler et al. from ORNL. The average SaltProc-calculated 5%-enriched uranium feed rate is 460.8 kg/y, which agrees well with the reference (480 kg/y). This dissertation illuminated the impact of xenon extraction efficiency on the long-term fuel cycle performance for the realistic reprocessing system model of the TAP concept with non-ideal removal efficiency. For limited gas removal efficiency, the fuel salt composition is strongly influenced by the neutron spectrum hardening due to the presence of neutron poisons (135Xe) in the core. Thus, more effective noble gas extraction significantly reduced neutron loss due to parasitic absorption, which led to better fuel utilization and extended core lifetime. Additionally, this work investigated MSR load-following capability through short-term depletion analysis with the power level variation (P) in [0,100%]. Online gas removal significantly improved the load-following capability of the MSBR by reducing xenon poisoning from -1457 pcm to -189 pcm. The TAP MSR demonstrated a negligible xenon poisoning effect even without online gas removal because its neutron energy spectrum is relatively fast throughout its lifetime. This work also analyzed safety parameter (temperature and void coefficient of reactivity, total control rod worth, kinetic parameters) variation during operation using fuel composition evolution obtained with SaltProc. On a lifetime-long timescale, the safety parameters worsened during operation for both considered MSRs due to a significant spectral shift. On a short-term timescale, the safety parameters during MSBR load-following slightly worsened right after power drop because 135Xe concentration peak caused substantial neutron spectrum hardening. However, during the next few hours, the gas removal system removed almost all 135Xe from the fuel, which led to significant improvement in all safety parameters. Overall, a reduced amount of neutron poisons (e.g., 135Xe) due to online gas extraction improved the safety case for both MSR designs. Finally, a simple uncertainty propagation via Monte Carlo depletion calculations in this work showed that the nuclear-data-related error (0.5-8% depending on the nuclide) is two orders of magnitude greater than the stochastic error (<0.07%).
Issue Date:2020-07-08
Type:Thesis
URI:http://hdl.handle.net/2142/108460
Rights Information:Copyright 2020 by Andrei Rykhlevskii.
Date Available in IDEALS:2020-10-07
Date Deposited:2020-08


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