2026-09-20T17:01:24Zhttps://dspace.mit.edu/server/oai/requestoai:dspace.mit.edu:1721.1/1139212022-01-13T07:54:05Zcom_1721.1_7582com_1721.1_7581col_1721.1_131023
Ronald G. Ballinger.
Lam, Stephen Tsz Tang
Massachusetts Institute of Technology. Department of Nuclear Science and Engineering.
Massachusetts Institute of Technology. Department of Nuclear Science and Engineering
2018-03-02T21:39:17Z
2018-03-02T21:39:17Z
2017
2017
http://hdl.handle.net/1721.1/113921
1023497861
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017.
This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 187-197).
The Fluoride Salt-Cooled High-Temperature Reactor (FHR) is an advanced reactor concept, that uses molten-salt coolant and solid-uranium fuel composed of graphite and silicon carbide-encapsulated tri-structural isotropic (TRISO) particles. The primary coolant salt is known as flibe (7Li2BeF4), which was chosen for its desirable thermal-hydraulic and neutronic properties. Under irradiation, coolant salts containing lithium capture neutrons generating tritium in quantities that are several orders of magnitude larger than the amounts generated by existing light water reactors. Adsorption technology is proposed, using chemically compatible carbon materials for the capture and control of tritium in the FHR. Various nanoporous activated carbon, graphene and nuclear graphite materials have been characterized. This includes the determination of BET surface area, total pore volume, average pore size, and pore size distribution by performing low-temperature gas adsorption experiments and applying microscopic thermodynamic theory. In addition, morphological analysis was conducted with scanning electron microscopy. Hydrogen was used as a surrogate. Its chemisorption on these materials have been measured and modeled at the reactor conditions of 700°C and pressures under 4 kPa. Models suggest that the total measured solubility of hydrogen includes a combination of dissociative and molecular adsorption. Carbon materials containing larger volumetric fractions of micropores (width < 2 nm) generally exhibited a higher hydrogen capacity. Further, the presence of micropores was associated with a relatively weak and reversible form of hydrogen chemisorption. At 500 Pa, microporous carbon materials captured 50 times more hydrogen than graphite, which was previously known to be the largest hydrogen sink at reactor conditions. The coupled effects of generation, chemical speciation, adsorption and diffusion of tritium in the FHR system were simulated over 200 full-power days. It was found that an adsorption column using high-performance carbon-based catalyst adsorbed substantial amounts of tritium and reduced the peak release rate from 2400 Ci/day to 40 Ci/day for the 236 MWt FHR. Further, the total tritium inventory in the system decreased by more than 70%, from 68,400 Ci to 19,400 Ci. This demonstrates that adsorption technology can greatly reduce the risk of radiological release during normal operation and reactor transient events.
by Stephen Tsz Tang Lam.
S.M.
215 pages
eng
Massachusetts Institute of Technology
MIT theses are protected by copyright. They may be viewed, downloaded, or printed from this source but further reproduction or distribution in any format is prohibited without written permission.
http://dspace.mit.edu/handle/1721.1/7582
Nuclear Science and Engineering.
Managing tritium inventory and release with carbon materials in a fluoride salt-cooled high-temperature reactor
Thesis
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<Abstract>The Fluoride Salt-Cooled High-Temperature Reactor (FHR) is an advanced reactor concept, that uses molten-salt coolant and solid-uranium fuel composed of graphite and silicon carbide-encapsulated tri-structural isotropic (TRISO) particles. The primary coolant salt is known as flibe (7Li2BeF4), which was chosen for its desirable thermal-hydraulic and neutronic properties. Under irradiation, coolant salts containing lithium capture neutrons generating tritium in quantities that are several orders of magnitude larger than the amounts generated by existing light water reactors. Adsorption technology is proposed, using chemically compatible carbon materials for the capture and control of tritium in the FHR. Various nanoporous activated carbon, graphene and nuclear graphite materials have been characterized. This includes the determination of BET surface area, total pore volume, average pore size, and pore size distribution by performing low-temperature gas adsorption experiments and applying microscopic thermodynamic theory. In addition, morphological analysis was conducted with scanning electron microscopy. Hydrogen was used as a surrogate. Its chemisorption on these materials have been measured and modeled at the reactor conditions of 700°C and pressures under 4 kPa. Models suggest that the total measured solubility of hydrogen includes a combination of dissociative and molecular adsorption. Carbon materials containing larger volumetric fractions of micropores (width < 2 nm) generally exhibited a higher hydrogen capacity. Further, the presence of micropores was associated with a relatively weak and reversible form of hydrogen chemisorption. At 500 Pa, microporous carbon materials captured 50 times more hydrogen than graphite, which was previously known to be the largest hydrogen sink at reactor conditions. The coupled effects of generation, chemical speciation, adsorption and diffusion of tritium in the FHR system were simulated over 200 full-power days. It was found that an adsorption column using high-performance carbon-based catalyst adsorbed substantial amounts of tritium and reduced the peak release rate from 2400 Ci/day to 40 Ci/day for the 236 MWt FHR. Further, the total tritium inventory in the system decreased by more than 70%, from 68,400 Ci to 19,400 Ci. This demonstrates that adsorption technology can greatly reduce the risk of radiological release during normal operation and reactor transient events.</Abstract>
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