{"id":{"repo_id":"ohiolink","oai_identifier":"oai:etd.ohiolink.edu:osu1364909574"},"canonical_url":"https://search.dev.ndltd.org/etd/ohiolink/oai:etd.ohiolink.edu:osu1364909574","repository":{"repo_id":"ohiolink","name":"OhioLINK","base_url":"https://etd.ohiolink.edu/acprod/odb_etd/ws/oai/oai"},"display":{"title":"Reconsideration of Inherent Neutron Sources in Liquid Fuel of Molten Salt Reactors","abstract":"Calculation of inherent neutron sources from alpha-n and gamma-n reactions in molten fluoride fuel salts using current gamma-n and alpha-n cross-sections and codes where possible. Original inherent neutron source was calculated for Molten Salt Reactor Experiment (MSRE) in 1959 using then-determined alpha-n cross-sections and an estimate for gamma-n cross-section for beryllium-9. Values selected to generate conservative neutron population value for determining required external neutron source and operational neutron population. Inherent neutron source values have not been revisited for molten salt fluid fuel reactors since. Updated alpha-n reaction rates per cubic centimeter calculated using SOURCES4C, compared against original Oak Ridge National Laboratory calculations for MSRE. Fuel salt aging simulated using ORIGEN for 300 days (no power) and a 150-day power cycle at 200 kWt followed by 150 days cooling, alpha-n source calculated by SOURCES4C and compared to alpha-n rates given by ORIGEN. Gamma-n reaction rates calculated from updated normalized cross-sections and compared to original ORNL 0.5 mb cross-section, which was for beryllium-9 salt component only. Recalculated reaction rates are markedly greater than original MSRE rates in fresh (unaged) fuel salt.","abstract_html":"Calculation of inherent neutron sources from alpha-n and gamma-n reactions in molten fluoride fuel salts using current gamma-n and alpha-n cross-sections and codes where possible. Original inherent neutron source was calculated for Molten Salt Reactor Experiment (MSRE) in 1959 using then-determined alpha-n cross-sections and an estimate for gamma-n cross-section for beryllium-9. Values selected to generate conservative neutron population value for determining required external neutron source and operational neutron population. Inherent neutron source values have not been revisited for molten salt fluid fuel reactors since. Updated alpha-n reaction rates per cubic centimeter calculated using SOURCES4C, compared against original Oak Ridge National Laboratory calculations for MSRE. Fuel salt aging simulated using ORIGEN for 300 days (no power) and a 150-day power cycle at 200 kWt followed by 150 days cooling, alpha-n source calculated by SOURCES4C and compared to alpha-n rates given by ORIGEN. Gamma-n reaction rates calculated from updated normalized cross-sections and compared to original ORNL 0.5 mb cross-section, which was for beryllium-9 salt component only. Recalculated reaction rates are markedly greater than original MSRE rates in fresh (unaged) fuel salt.","abstract_has_math":false,"creators":["Powell, Walter Newton"],"institution":"The Ohio State University","degree_name":"Master of Science","degree_level":"masters","degree_discipline":"Nuclear Engineering","degree_department":null,"school":null,"contributors":["Cao, Lei"],"advisors":[],"committee_chairs":[],"committee_members":[],"year":2013,"date_issued":"2013-07-05","date_published":"2013-07-05","updated_at":"2026-07-24T03:37:16Z","subjects":["Nuclear Engineering","reactor fuels","fuel salts","alpha-n","gamma-n","neutron sources","molten salt reactor"],"languages":["English"],"rights":["unrestricted","This thesis or dissertation is protected by copyright: all rights reserved. It may not be copied or redistributed beyond the terms of applicable copyright laws."],"rights_urls":[],"identifier_entries":[]},"links":{"outbound_url":"http://rave.ohiolink.edu/etdc/view?acc_num=osu1364909574","outbound_label":"Repository record","outbound_source":"dc:identifier"},"metadata_groups":[{"id":"people","label":"People","entries":[{"key":"dc:contributor","label":"Contributor","values":["Cao, Lei"]},{"key":"dc:creator","label":"Author","values":["Powell, Walter Newton"]}]},{"id":"academic_context","label":"Academic Context","entries":[{"key":"dc:date","label":"Dc Date","values":["2013-07-05"]},{"key":"dc:publisher","label":"Institution","values":["The Ohio State University / OhioLINK"]},{"key":"dc:type","label":"Dc Type","values":["Electronic Thesis or Dissertation"]},{"key":"thesis:degree_discipline","label":"Discipline","values":["Nuclear Engineering"]},{"key":"thesis:degree_level","label":"Degree Level","values":["masters"]},{"key":"thesis:degree_name","label":"Degree Name","values":["Master of Science"]},{"key":"thesis:institution_name","label":"Thesis Institution Name","values":["The Ohio State University"]}]},{"id":"subjects_keywords","label":"Subjects and Keywords","entries":[{"key":"dc:subject","label":"Dc Subject","values":["Nuclear Engineering","reactor fuels","fuel salts","alpha-n","gamma-n","neutron sources","molten salt reactor"]}]},{"id":"language_rights","label":"Language and Rights","entries":[{"key":"dc:language","label":"Dc Language","values":["English"]},{"key":"dc:rights","label":"Dc Rights","values":["unrestricted","This thesis or dissertation is protected by copyright: all rights reserved. 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Inherent neutron source values have not been revisited for molten salt fluid fuel reactors since. Updated alpha-n reaction rates per cubic centimeter calculated using SOURCES4C, compared against original Oak Ridge National Laboratory calculations for MSRE. Fuel salt aging simulated using ORIGEN for 300 days (no power) and a 150-day power cycle at 200 kWt followed by 150 days cooling, alpha-n source calculated by SOURCES4C and compared to alpha-n rates given by ORIGEN. Gamma-n reaction rates calculated from updated normalized cross-sections and compared to original ORNL 0.5 mb cross-section, which was for beryllium-9 salt component only. Recalculated reaction rates are markedly greater than original MSRE rates in fresh (unaged) fuel salt."]},{"key":"dc:format","label":"Dc Format","values":["application/pdf","p.111","1.15 MB"]},{"key":"dc:title","label":"Title","values":["Reconsideration of Inherent Neutron Sources in Liquid Fuel of Molten Salt Reactors"]}]}],"canonical_facts":{"dc:contributor":["Cao, Lei"],"dc:creator":["Powell, Walter Newton"],"dc:date":["2013-07-05"],"dc:description":["Calculation of inherent neutron sources from alpha-n and gamma-n reactions in molten fluoride fuel salts using current gamma-n and alpha-n cross-sections and codes where possible. Original inherent neutron source was calculated for Molten Salt Reactor Experiment (MSRE) in 1959 using then-determined alpha-n cross-sections and an estimate for gamma-n cross-section for beryllium-9. Values selected to generate conservative neutron population value for determining required external neutron source and operational neutron population. Inherent neutron source values have not been revisited for molten salt fluid fuel reactors since. Updated alpha-n reaction rates per cubic centimeter calculated using SOURCES4C, compared against original Oak Ridge National Laboratory calculations for MSRE. Fuel salt aging simulated using ORIGEN for 300 days (no power) and a 150-day power cycle at 200 kWt followed by 150 days cooling, alpha-n source calculated by SOURCES4C and compared to alpha-n rates given by ORIGEN. Gamma-n reaction rates calculated from updated normalized cross-sections and compared to original ORNL 0.5 mb cross-section, which was for beryllium-9 salt component only. Recalculated reaction rates are markedly greater than original MSRE rates in fresh (unaged) fuel salt."],"dc:format":["application/pdf","p.111","1.15 MB"],"dc:identifier":["http://rave.ohiolink.edu/etdc/view?acc_num=osu1364909574"],"dc:language":["English"],"dc:publisher":["The Ohio State University / OhioLINK"],"dc:rights":["unrestricted","This thesis or dissertation is protected by copyright: all rights reserved. It may not be copied or redistributed beyond the terms of applicable copyright laws."],"dc:subject":["Nuclear Engineering","reactor fuels","fuel salts","alpha-n","gamma-n","neutron sources","molten salt reactor"],"dc:title":["Reconsideration of Inherent Neutron Sources in Liquid Fuel of Molten Salt Reactors"],"dc:type":["Electronic Thesis or Dissertation"],"thesis:degree_discipline":["Nuclear Engineering"],"thesis:degree_level":["masters"],"thesis:degree_name":["Master of Science"],"thesis:institution_name":["The Ohio State University"]},"updated_at":"2026-07-24T03:37:16Z"}