02272oam 2200541I 450 991070741270332120170719110526.0(CKB)5470000002465405(OCoLC)957772481(OCoLC)995470000002465405(EXLCZ)99547000000246540520160906j201608 ua 0engur|||||||||||txtrdacontentcrdamediacrrdacarrierFuel rod behavior and uncertainty analysis by FRAPTRAN/TRACE/DAKOTA code in maanshan LBLOCA /prepared by Chunkuan Shih [and five others]Washington, DC:Division of Systems Analysis, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,August 2016.1 online resource (various pagings) illustrationsInternational agreement report ;NUREG/IA-0471"Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation.""Department of Nuclear Safety, Taiwan Power Company.""K. Tien, NRC project manager.""Manuscript completed: March 2016; date published: August 2016.""Prepared as part of the agreement on research participation and technical exchange under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)."Includes bibliographical references.Nuclear fuel rodsNuclear reactor accidentsAnalysisPressurized water reactorsLoss of coolantAnalysisNuclear power plantsTaiwanNuclear fuel rods.Nuclear reactor accidentsAnalysis.Pressurized water reactorsLoss of coolantAnalysis.Nuclear power plantsShih Chunkuan1389487U.S. Nuclear Regulatory Commission.Division of Systems Analysis,GPOGPOGPONKFCUTTFWGPOBOOK9910707412703321Fuel rod behavior and uncertainty analysis by FRAPTRAN3479526UNINA