Integrity control of an RBMK-1500 fuel rod locally oxidized under a bounding reactivity-initiated accident
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摘要
In 2007, the license for the second reactor unit of the Ignalina nuclear power plant was renewed considering the safety-related modifications introduced in this reactor.The Safety Analysis Report for this reactor unit was prepared with more strict criteria. The bounding reactivityinitiated accident(RIA) performed by the Lithuanian Energy Institute could be mentioned as an example. The performed analysis demonstrated that even when the worst initial conditions and possible uncertainties are considered,the fuel cladding remains intact. However, the analysis was performed assuming a fresh fuel assembly. In this study, an analysis of the fuel rod cladding behavior in the RBMK-1500 reactor following a bounding RIA is performed using the computational codes FEMAXI-6 and RELAP5. The analysis is extended by modeling an oxide layer(nodular corrosion) on the external surface cladding. An uncertainty and sensitivity analysis was performed using a method developed by the Society for Plant and Reactor Safety,employing the Software for Uncertainty and Sensitivity Analyses, in order to evaluate the effect of the oxide layer on the inside and outside fuel rod temperatures. The results of the thermo-mechanical analysis(stress, strain, and enthalpy) for a local oxide layer with a thickness of 70μm show that despite the exceeded limit of allowed linear power density, the fuel rod is under acceptable safety conditions.
In 2007, the license for the second reactor unit of the Ignalina nuclear power plant was renewed considering the safety-related modifications introduced in this reactor.The Safety Analysis Report for this reactor unit was prepared with more strict criteria. The bounding reactivityinitiated accident(RIA) performed by the Lithuanian Energy Institute could be mentioned as an example. The performed analysis demonstrated that even when the worst initial conditions and possible uncertainties are considered,the fuel cladding remains intact. However, the analysis was performed assuming a fresh fuel assembly. In this study, an analysis of the fuel rod cladding behavior in the RBMK-1500 reactor following a bounding RIA is performed using the computational codes FEMAXI-6 and RELAP5. The analysis is extended by modeling an oxide layer(nodular corrosion) on the external surface cladding. An uncertainty and sensitivity analysis was performed using a method developed by the Society for Plant and Reactor Safety,employing the Software for Uncertainty and Sensitivity Analyses, in order to evaluate the effect of the oxide layer on the inside and outside fuel rod temperatures. The results of the thermo-mechanical analysis(stress, strain, and enthalpy) for a local oxide layer with a thickness of 70μm show that despite the exceeded limit of allowed linear power density, the fuel rod is under acceptable safety conditions.
In 2007, the license for the second reactor unit of the Ignalina nuclear power plant was renewed considering the safety-related modifications introduced in this reactor.The Safety Analysis Report for this reactor unit was prepared with more strict criteria. The bounding reactivityinitiated accident(RIA) performed by the Lithuanian Energy Institute could be mentioned as an example. The performed analysis demonstrated that even when the worst initial conditions and possible uncertainties are considered,the fuel cladding remains intact. However, the analysis was performed assuming a fresh fuel assembly. In this study, an analysis of the fuel rod cladding behavior in the RBMK-1500 reactor following a bounding RIA is performed using the computational codes FEMAXI-6 and RELAP5. The analysis is extended by modeling an oxide layer(nodular corrosion) on the external surface cladding. An uncertainty and sensitivity analysis was performed using a method developed by the Society for Plant and Reactor Safety,employing the Software for Uncertainty and Sensitivity Analyses, in order to evaluate the effect of the oxide layer on the inside and outside fuel rod temperatures. The results of the thermo-mechanical analysis(stress, strain, and enthalpy) for a local oxide layer with a thickness of 70μm show that despite the exceeded limit of allowed linear power density, the fuel rod is under acceptable safety conditions.
引文
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2.R.Pabarcius,A.Kaliatka,A.Marao,Analysis of fuel rod behaviour during limiting RIA in RBMK plants.J.Kerntechnik 75,329-336(2010)
3.IAEA,Deterministic Safety Analysis for Nuclear Power Plants.Specific Safety Guide No.SSG-2(Vienna,2009)
4.QUABOX-CUBBOX software,https://www.grs.de/en/simula tion-codes/quabox-cubbox
5.M.Suzuki,H.Saitou,Light Water Reactor Fuel Analysis Code FEMAXI-6(Ver.1),detailed structure and user’s Manual.JAEA-Data/Code 2005-003(2006)
6.C.D.Fletcher,R.R.Schult,RELAP5/MOD3 Code Manual,User’s Guidelines,NUREG/CR-5535,INEL-95/0174,Vol.5,Rev.1(1995)
7.IAEA,Accident analysis for nuclear power plants with graphite moderated boiling water RBMK reactors.Safety Reports Series No.43(Vienna,2005)
8.R.Pabarcˇius,A.Tonkuˉnas,Definition of conservative conditions for RIA analysis in the modernized RBMK reactor core.J.Kerntechnik 75,178-184(2010)
9.L.Lunde,Special features of external corrosion of fuel cladding in boiling water reactors.Nucl.Eng.Des.33,178-195(1975).https://doi.org/10.1016/0029-5493(75)90021-7
10.A.Marao,T.Kaliatka,E.Usˇpuras,Analysis of processes in RBMK-1500 fuel assemblies during normal operation cycle.J.ENERGETIKA 56,8-18(2010)
11.W.G.Luscher,K.J.Geelhood,Material Property Correlations:Comparisons between FRAPCON-3.5,FRAPTRAN-1.5,and MATPRO.NUREG/CR-7024 Rev.1 PNNL-19417 Rev.1(2014)
12.D.Gomes,A.Teixeira,Angra1 high burnup fuel behavior under reactivity initiated accident conditions.In:2011 International Nuclear Atlantic Conference.Belo Horizonte,MG,Brazil,October 24-28(2011)
13.C.Vitanza,U.Graziani,N.T.Fordestrommen,K.O.Vilpponen,Fission Gas Release from in-pile measurements.Report HPR-221.10,OECD Halden Reactor Project,(1978).
14.C.Vitanza,E.Kolstad,C.Graziani,Fission gas release from UO2 pellet at high burnup,Topical Meeting on Light Water Reactor Fuel Performance(American NuclearSociety,Portland,1979)
15.Organisation for Economic Co-Operation and DevelopmentNuclear Energy Agency,Nuclear Fuel Safety Criteria Technical Review,64 p Paris(France)(2001)
16.H.Glaser,GRS Method for Uncertainty and Sensitivity Evaluation of Code Results and Applications.Sci.Tech.of Nucl.Installations,Vol.2008,Article ID 798901,p 7(2008).https://doi.org/10.1155/2008/798901
17.M.Kloos,E.Hofer,SUSA Version 3.5.User’s guide and tutorial.GRS,Garching,(2002)
18.K.J.Geelhood,W.G.Luscher,C.E.Beyer and al.Predictive bias and Sensitivity in NRC fuel performance codes.NUREG/CR-7001,PNNL-17644(2009).NRC Job Code N6326
19.D.L.Hagrman,G.A.Reymann,MATPRO-version 11.Ahandbook of materials properties for use in the analysis of light water reactor fuel rod behaviour.Technical report,NUREG/CR-0497,TREE-1280(1979)
20.IAEA-TECDOC-1496,Thermophysical properties database of materials for light water reactors and heavy water reactors.Final report of a coordinated research project(1999-2005)