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dc.contributor.authorHeising, Carolyn D. (Carolyn DeLane), 1952-en_US
dc.contributor.authorLepervanche-Valencia, José Gregorioen_US
dc.contributor.authorPilat, E. E., 1937-en_US
dc.contributor.authorSlifer, Bruce C.en_US
dc.contributor.otherMassachusetts Institute of Technology. Department of Nuclear Engineeringen_US
dc.contributor.otherYankee Atomic Electric Companyen_US
dc.date.accessioned2014-09-16T23:37:30Z
dc.date.available2014-09-16T23:37:30Z
dc.date.issued1980en_US
dc.identifier.urihttp://hdl.handle.net/1721.1/89741
dc.descriptionIncludes bibliographical referencesen_US
dc.descriptionFinal Report; July 1980en_US
dc.description.abstractPost-accident hydrogen generation in BWR containments is analyzed as a function of engineered hydrogen control system, assumed either nitrogen inerting or air dilution. Fault tree analysis was applied to assess the failure probability per demand of each system. These failure rates were then combined with the probability of accidents producing various hydrogen generation rates to calculate the overall system hydrogen control probability. Results indicate that both systems render approximately the same overall hydrogen control probability (air dilution: .917 - .989; nitrogen inerting: .987 - .998). Drywell entries and unscheduled shutdowns were also analyzed to determine the impact on the total BWR accident risk as it relates to the decay heat removal system. Results indicate that inerting may increase the overall risk due to a possible increase in the number of unscheduled shutdowns due to a lessened operator ability to correct and identify "unidentified" leakage from the primary coolant system. Further, possible benefits of inerting due to reduced torus corrosion and fire risk in containment appear to be dominated by the possible operations related disbenefits.en_US
dc.format.extent123, A1-A9 pagesen_US
dc.publisherCambridge, Mass. : Massachusetts Institute of Technology, [Dept. of Nuclear Engineering, 1980]en_US
dc.relation.ispartofseriesMITNE ; no. 240en_US
dc.subject.lccTK9008.M41 N96 no.240en_US
dc.subject.lcshVermont Yankee Nuclear Power Station (Vernon, Vt.)en_US
dc.subject.lcshBoiling water reactors -- Coolingen_US
dc.subject.lcshBoiling water reactors -- Safety measuresen_US
dc.titleAnalyzing the safety impact of containment inerting at Vermont Yankeeen_US
dc.typeTechnical Reporten_US
dc.identifier.oclc857467927en_US


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