ML11308A493

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Grand Gulf, Applicants Environmental Report, Operating License Renewal Stage, Attachment E - Severe Accident Mitigation Alternatives Analysis
ML11308A493
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 10/28/2011
From:
Entergy Operations
To:
Office of Nuclear Reactor Regulation
References
GNRO-2011/00093
Download: ML11308A493 (159)
v • d • e


Text

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E-1Attachment ESevere Accident Mitigation Alternatives AnalysisAttachment E contains the following sections.E.1 - Evaluation of GGNS PSA ModelE.2 - Evaluation of GGNS SAMA Candidates Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage iTABLE OF CONTENTSE.1EVALUATION OF PROBABILISTIC SAFETY ANALYSIS MODEL. . . . . . . . . . . . . .E.1-1E.1.1PSA Model - Level 1 Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-1E.1.2PSA Model - Level 2 Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-23E.1.2.1Containment Performance Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-23E.1.2.2Radionuclide Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-48E.1.2.2.1Introduction. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-48E.1.2.2.2Timing of Release. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-48 E.1.2.2.3Magnitude of Release. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-48E.1.2.2.4Release Category Bin Assignments. . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-49E.1.2.2.5Mapping of Level 1 Results into the Various Release Categories . . . . . .E.1-50 E.1.2.2.6Process Used to Group the Source Terms. . . . . . . . . . . . . . . . . . . . . . . .E.1-54E.1.2.2.7Consequence Analysis Source Terms . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-56E.1.2.2.8Release Magnitude Calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-56E.1.3IPEEE Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E

.1-59E.1.3.1Seismic Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-59E.1.3.2Fire Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-59E.1.3.3Other External Hazards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-67E.1.4PSA Model Revisions and Peer Review Summary . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-68E.1.4.1Major Differences between the 1997(R1) PSA Model and the IPE Model. . . . .E.1-68E.1.4.2Major Differences between the 2002 (R2) PSA Model and the 1997(R1) PSA Model. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-68E.1.4.3Major Differences between the 2010 (R3) PSA Model and the 2002 (R2) PSA Model. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-69E.1.4.4Major Differences between the 2010 (EPU) PSA Model and the 2010 (R3) PSA Model. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-69E.1.4.5PSA Model Peer Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-71E.1.5The MACCS2 Model-Level 3 Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-71E.1.5.1Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-71E.1.5.2Input. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

E.1-72E.1.5.2.1Projected Total Population by Spatial Element. . . . . . . . . . . . . . . . . . . . .E.1-72E.1.5.2.2Land Fraction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-74E.1.5.2.3Watershed Class . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-74 E.1.5.2.4Regional Economic Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-74E.1.5.2.5Agriculture Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-75E.1.5.2.6Meteorological Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-75 E.1.5.2.7Emergency Response Assumptions. . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-76E.1.5.2.8Core Inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-76E.1.5.2.9Source Terms. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-78E.1.5.3RESULTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-78E.1.6References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-81 Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage iiE.2EVALUATION OF GGNS SAMA CANDIDATES. . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.2-1E.2.1SAMA List Compilation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.2-1E.2.2Qualitative Screening of SAMA Candidates (Phase I). . . . . . . . . . . . . . . . . . . . . . . .E.2-2E.2.3Final Screening and Cost Benefit Evaluation of SAMA Candidates (Phase II) . . . . .E.2-2E.2.4Sensitivity Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.

2-15E.2.5References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.2-17 Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal StageiiiLIST OF TABLESTable E.1-1GGNS EPU Model CDF Results by Major Initiators . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-3Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF) . . E.1-5Table E.1-3Notation and Definitions for GGNS CET Functional Nodes Description. . . . . . . . . . . . . E.1-25Table E.1-4Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-40Table E.1-6GGNS Release Categories. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-50Table E.1-5Release Severity and Timing Classification Scheme Summary. . . . . . . . . . . . . . . . . . . E.1-50Table E.1-7Summary of GGNS Core Damage Accident Sequences Plant Damage States. . . . . . . E.1-52Table E.1-8Summary of Containment Event Tree Quantification . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-55Table E.1-9GGNS Release Category Source Terms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-57Table E.1-10GGNS Fire IPEEE Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-61Table E.1-11Estimated Population Distribution within a 50-Mile Radius. . . . . . . . . . . . . . . . . . . . . . . E.1-73Table E.1-12Estimated GGNS Core Inventory (Becquerels). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-77Table E.1-13Base Case Mean PDR and OECR Values for Postulated Internal Events. . . . . . . . . . . E.1-79Table E.1-14Summary of Offsite Consequence Results for Sensitivity Results . . . . . . . . . . . . . . . . . E.1-80 Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage ivTable E.2-1Phase I SAMAs Related to IPE and IPEEE Insights. . . . . . . . . . . . . . . . . . . . . . . . . . . . E.2-19Table E.2-2Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation . . . . . E.2-30Table E.2-3Sensitivity Analysis Results. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.2-5 7

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage v LIST OF FIGURESFigure E.1-1GGNS Radionuclide Release Category Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-38Figure E.1-2Summary of GGNS Core Damage Accident Sequences Plant Damage States. . . . . . . E.1-39 Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage viLIST OF ACRONYMS IN ATTACHMENTS E.1 AND E.2 Acronym DefinitionACAlternating currentADSAutomatic depressurization system ASDSAlternate shutdown system ATWSAnticipated transient without scram BWRBoiling water reactor BWROGBoiling Water Reactor Owners Group CCFCommon cause failure CCWComponent cooling water CDFCore damage frequency CETContainment event tree CNSCooper Nuclear Station CPIConsumer Price Index CRDControl rod drive CsICesium iodide CSTCondensate storage tank DCDirect current DFDecontamination factor DGDiesel generator ECCSEmergency core cooling system EDGEmergency diesel generator EOPEmergency operating procedure EPGEmergency procedure guidelines EPUExtended power uprate EPZEmergency planning zone Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage viiESFEmergency safety featureFIVEFire induced vulnerability evaluation FPSFire protection system FPWFire protection water FWFeedwater GGNSGrand Gulf Nuclear Station HPCIHigh pressure coolant injection HPCSHigh pressure core spray HVACHeating, ventilation and air conditioningIAInstrument air IPEIndividual Plant Examination IPEEEIndividual Plant Examination of External EventsISLOCAInterfacing systems loss of coolant accident LERFLarge early release frequency LOCALoss of coolant accident LOSPLoss of off-site power LPCILow pressure coolant injection LPCSLow pressure core spray MAAPModular Accident Analysis Program MACCS2MELCOR Accident Consequences Code System 2 MSIVMain steam isolation valve MSTMain steam tunnel NRCNuclear Regulatory Commission OECROff-site economic cost risk OSPOff-site power PCPLPrimary containment pressure limit Acronym Definition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage viiiPCSPower conversion systemPDRPopulation dose risk PDSPlant damage state PHVPump house ventilation PRAProbabilistic Risk Assessment PSAProbabilistic Safety Assessment PSWPlant service water RBReactor building RCICReactor core isolation cooling RCSReactor coolant system RHRResidual heat removal RPVReactor pressure vessel RRWRisk reduction worth SAGSevere accident guideline SAMA Severe accident mitigation alternative SBOStation blackout SLCStandby liquid control SORVStuck open relief valve SPCSuppression pool cooling SPMUSuppression pool makeup SRVSafety relief valve SSWStandby service water TBCWTurbine building cooling water WWWetwell Acronym Definition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal StageATTACHMENT E.1 EVALUATION OF GGNS PSA MODEL Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-1E.1 EVALUATION OF PROBABILISTIC SAFETY ANALYSIS MODELThe severe accident risk was estimated using the Probabilistic Safety Analysis (PSA) model and a Level 3 model developed using version 1.13.1 of the MELCOR Accident Consequences Code System version 2 (MACCS2 code). The CAFTA code was used to develop the Grand Gulf Nuclear Station (GGNS) PSA Level 1 and Level 2 models. This section provides the description of GGNS PSA levels 1, 2, and 3 analyses, Core Damage Frequency (CDF) uncertainty, Individual Plant Examination of External Events (IPEEE) analyses, and PSA model peer review.E.1.1PSA Model - Level 1 AnalysisThe PSA model (Level 1 and Level 2) used for the Severe Accident Mitigation Alternative (SAMA) analysis was the most recent internal events risk model, reflecting the GGNS extended power uprate (EPU) configuration [

E.1-18 and E.1-4]. In the EPU model, the Rev. 3 model which reflects GGNS design, component failure and unavailability data as of August 2006 was modified to reflect the EPU configuration. There have been no major plant hardware changes or procedural modifications since August 2006 that would have a significant impact on the results of the SAMA analysis. Thus, the EPU model used for the SAMA analysis is appropriate. The GGNS model adopts the small event tree / large fault tree approach and uses the CAFTA code for quantifying CDF.The PSA model has had three major revisions since the Individual Plant Examination (IPE) due to the following.*Equipment performance: As data collection progresses, estimated failure rates and system unavailability data change.*Plant configuration changes: Plant configuration changes are incorporated into the PSA model.*Modeling changes: The PSA model is refined to incorporate the latest state of knowledge and recommendations from internal and industry peer reviews.In the EPU model, the Rev. 3 model was modified to reflect the EPU configuration. The EPU model contains the major initiators leading to core damage with baseline CDFs listed in TableE.1-1

.The GGNS L1 Model was reviewed to identify those potential risk contributors that made a significant contribution to CDF. CDF-based Risk Reduction Worth (RRW) rankings were reviewed down to 1.005. Events below this point would influence the CDF by less than 0.5percent and are judged to be highly unlikely contributors for the identification of cost-beneficial enhancements. These basic events-including component failures, operator actions, and initiating events-were reviewed to determine if additional SAMA actions may need to be considered.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-2TableE.1-2 provides a correlation between the Level 1 RRW risk significant events (component failures, operator actions, and initiating events) down to 1.005 identified from the GGNS PSA model and the SAMAs evaluated in Section E.2.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-3Table E.1-1GGNS EPU Model CDF Results by Major InitiatorsInitiating Event GroupTotal IE Group Probability% CDFLarge Loss of Coolant Accident (LOCA)1.45E-077.10Feedwater Line Break Outside of Containment2.76E-100.00Plant Service Water (PSW) Flooding Initiator1.00E-090.00Reactor Vessel Rupture1.00E-080.50Intermediate LOCA2.03E-081.00Small LOCA1.33E-110.00Small-Small LOCA2.47E-110.00 Standby Service Water (SSW) Flooding Initiator6.55E-120.00Loss of Off-Site Power Initiator2.87E-0714.00Loss of 500 kV Power (Preferred)(1)5.12E-110.00Loss of Power Conversion System (PCS) Initiator2.31E-0711.20Closure of Main Steam Isolation Valves (MSIVs) (Initiator) 8.81E-084.30PCS Available Transient6.32E-0730.80Loss of Condensate Feed Water Pumps2.20E-0710.70Inadvertent Open Relief Valve9.78E-090.50Loss of Alternating Current (AC) Division 1 Initiator1.79E-080.90Loss of AC Division 2 Initiator3.82E-081.90Loss of Turbine Cooling Water (TBCW)8.09E-090.40Loss of Component Cooling Water Initiating Event6.87E-100.00Loss of Control Rod Drive (CRD)2.20E-090.10 Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-4Loss of Direct Current (DC) Division 1 Initiator2.22E-100.00Loss of DC Division 2 Initiator1.30E-100.00Loss of Instrument Air1.36E-076.60Loss of PSW Initiating Event1.50E-090.10 Loss of Service Transformer 119.20E-084.50Loss of Service Transformer 211.09E-075.30Interfacing System Loss of Coolant Accident (ISLOCA) in Shutdown Cooling Supply Header (Pen 14)2.03E-100.00Total CDF2.05E-06100.00Total Anticipated Transient without Scram (ATWS)(2)~ 3.08E-090.15Total Station Blackout (SBO)(2) (TB)~ 7.51E-0736.651.Loss of all 500 kV lines (preferred offsite power), for which the 115 kV line is still available to power the Emergency Safety Feature (ESF) loads following manual realignment of the vital buses.2.SBO and ATWS may occur following multiple initiators; thus their contributions to CDF are listed separately.Table E.1-1 (Continued)GGNS EPU Model CDF Results by Major InitiatorsInitiating Event GroupTotal IE Group Probability% CDF Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-5Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition%A3.19E-051.036Large LOCAThis term represents a large LOCA. Phase II SAMA 56 for detecting LOCAs was evaluated.%S14.69E-061.005Intermediate LOCAThis term represents an intermediate LOCA. Phase II SAMA 57 for implementing a GRA was evaluated.%T12.48E-021.6289Loss of offsite power initiatorThis term represents a loss of offsite power initiator. Phase II SAMAs 7, 15 and 18 for improving offsite, switchyard and transformer availability were evaluated.%T21.77E-011.0969Loss of PCS initiatorThis term represents a loss of power conversion system initiator. Phase II SAMA 28 for improving availability and reliability of feedwa ter was evaluated.%T2M2.01E-011.0423Closure of MSIVs (initiator)This term represents an inadvertent MSIV closure initiator. Phase II SAMAs 23, 37, 49, and 53 to improve SRV and MSIV availability and reliability and to reduce initiating event frequencies by implementing generation risk assessment were evaluated.%T3A7.98E-011.2633PCS available transientThis term represents a general initiator with PCS available. Phase II SAMA 57 for scram reduction modeling, and SAMAs 34, 35, 36, and 37 for improving instrument air reliability were evaluated.%T3B2.00E-011.0934Loss of condensate feed water pumpsThis term represents a loss of condensate feedwater pumps initiator. Phase II SAMA 28 for improving availability and reliability of feedwa ter was evaluated.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-6%TAC12.56E-031.0078Loss of AC Division 1 initiatorThis term represents the loss of AC Division 1 initiator. Phase II SAMAs 5 and 8 for enhancing AC system reliability or to cope with loss of offsite power and SBO events were evaluated.%TAC22.56E-031.0191Loss of AC Division 2 initiatorThis term represents the loss of AC Division 1 initiator. Phase II SAMAs 5 and 8 for enhancing AC system reliability or to cope with loss of offsite power and SBO events were evaluated.%TIA3.51E-031.0487Loss of instrument airThis term represents a loss of instrument air initiator. Phase II SAMAs 34, 35, 36, and 37 for improving the instrument air system were evaluated.%TST119.85E-021.029Loss of service transformer 11This term represents a loss of service transformer 11. Phase II SAMA 18 for protecting transformers was evaluated.%TST217.48E-021.0384Loss of service transformer 21This term represents a loss of service transformer 21. Phase II SAMA 18 for protecting transformers was evaluated.

B21-FO-HEBOTTLES1.00E+001.0629Operator fails to connect gas bottles to ADS air headerThis term represents a failure to manually operate ADS when IA is lost. Phase II SAMA 36 for adding automatic nitrogen backup to ADS components was evaluated.B21-FO-HEDEP2-I1.00E+001.5587Operator fails to manually depressurize vessel with non-ADS valvesThis term represents a failure to manually operate ADS when IA is lost. Phase II SAMAs 20, 21, 22, and 28 for improving high pressure injection or ADS components were evaluated.E12-CF-MVLPCS5.73E-051.0072Two or more LPSI and LPCS injection MOVs to openThis term represents a failure of LPCI injection valves to open. Phase II SAMA 25 for bypassing LPCI low pressure permissives was evaluated.E12-LF-FGCS1.00E+001.0637Containment spray signal generatedThis term is a flag. No SAMAs need to be aligned.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-7E12-MA-TMRHRA8.83E-031.0118RHR Train A unavailable due to maintenanceThis term represents a failure of LPCI. Phase II SAMAs 24, and 25 for improving or adding low pressure injection systems were evaluated.E12-MA-TMRHRB5.56E-031.0066RHR Train B unavailable due to maintenanceThis term represents a failure of LPCI. Phase II SAMAs 24, and 25 for improving or adding low pressure injection systems were evaluated.E22-042-H6.40E-031.0377Suppression pool suction line hardware failure (long term)This term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.E22-CC-MVF004-G6.30E-031.0742Normally closed motor driven valve FOO4 fails to openThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.E22-CC-MVF012-G6.30E-031.0742Minimum flow valve F012-C fails to openThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.E22-FS-MPC001-G3.00E-031.0336HPCS motor driven pump C001 fails to startThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.E22-HW-ICHPCS-G1.60E-031.0174HPCS actuation circuit failureThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated. E22-HW-ICMNFLO-G1.60E-031.0174Minimum flow control circuit failureThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-8E22-MA-MAHPCS-G6.59E-031.0662HPCS unavailable due to maintenanceThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.E22-OO-MVF012-G3.40E-031.0384Normally open motor driven valve E22-F012 fails to closeThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.E51-021-H7.99E-031.0127Suppression pool suction switchover fails due to hardware (long term)This term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.E51-026-G6.40E-031.0634RCIC pump fails-minimum flow path fails to openThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.E51-035M-G1.27E-021.1348RCIC steam supply valves failThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.E51-043-G8.29E-031.0839Lube oil cooling line hardware failureThis term represents a failure of a high pressure injection. Phase II SAMA 63 for improving RCIC reliability was evaluated. E51-CC-MVF013A-G6.30E-031.0623Motor-operated valve F013-A fails to openThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.E51-FF-FSC001-G3.51E-031.0335RCIC pump start failuresThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-9 E51-FO-HEF031A-G1.00E+001.0063Operator fails to open SP suction valve F031-AThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E51-FO-HETRPBYP1.00E+001.0307Human error: Failure to bypass RCIC temperature trips (EOP Attachment 3)This term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.E51-FR-TPC001-G2.01E-011.2014RCIC turbine-driven pump fails to runThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E51-FR-TPC18HR-G6.71E-021.0586RCIC turbine fails to run for 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />sThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.E51-HW-ICLVL8-I1.60E-031.0148Hardware failure of level 8 isolation circuitryThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.E51-HW-ICSYACT-G1.60E-031.0148RCIC actuation circuitry failureThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.E51-MA-TMRCIC-G1.24E-021.1075RCIC unavailable due to maintenanceThis term represents a failure of a high pressure injection. Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

HVC-LF-FGSSWAPH1.00E+001.0113Failure of SSW A pump house ventilationThis term is a flag. No SAMAs need to be aligned.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-10L21-OP-BT-1A3-D1.00E+001.0382Battery 1A3 discharged (~8hours depletion time)This term represents battery depletion before recovery of offsite power. Phase II SAMAs 1, 2, 3, 11,12, and 15 for extending available recovery time by improving DC power were evaluated.LOSP-EPRI1.00E-031.022Conditional LOSP after a plant tripThis term represents a transient induced loss of offsite power. Phase II SAMAs 5 and 8 for improving AC power reliability were evaluated.M24-RP-CTFLECCS9.38E-031.0714ECCS pump failure due to containment failureThis term represents loss of ECCS equipment due to containment failure. Phase II SAMAs 20, 28, 39, 41, and 60 for adding or improving injection sources not affected by a containment failure and SAMAs 19, 46 and 47 for improving the reliability of the containment vent were evaluated. M41-FF-MLVNTHW-Q7.98E-031.0099Hardware failure of the containment venting valvesThis term represents a failure of the containment vent valves. Phase II SAMAs 38, 39, 40, 41, 42, 46 and 47 for providing better suppression pool cooling, containment spray and a passive containment vent were evaluated.N21-FO-HELVL9-I1.00E+001.0827Human error: Failure to restart reactor feed pumps following level 9 tripThis term represents a failure of a human action to restore feedwater and manually depressurize. Phase II SAMAs 20 and 61 for improving high pressure injection capability were evaluated.N21-FO-HEPCS-G1.00E+001.1081Human error: Failure to properly align the PCS for injectionThis term represents a failure of a human action to restore feedwater and manually depressurize. Phase II SAMAs 20 and 40 for improving high pressure injection and suppression pool cooling capability were evaluated.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal StageE.1-11NR-ACHWR-1HRS6.00E-011.011Failure to recover AC bus failure in 1 hourThis term represents a failure to recover the AC bus. Phase II SAMAs 5, 8, 17 and 18 for protecting or providing alternate bus power supplies were evaluated.NR-ACHWR-8HRS1.00E-021.0158Failure to recover AC bus failure in 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />sThis term represents a failure to recover the AC bus. Phase II SAMAs 5, 8, 17 and 18 for protecting or providing alternate bus power supplies were evaluated.NRC-DEP-RCIC8.40E-031.0051Failure to manually depressurize using RCICThis term represents a failure of a human action to manually depressurize using RCIC. Phase II SAMA 22 increased ADS reliability was evaluated.NRC-DG-CF1HRS9.00E-011.0059Failure to recover diesel generator common cause failure in 1 hourThis term represents a failure of a human action to recover the DG common cause failure in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Phase II SAMAs 5 and 8 to install an additional diesel or gas turbine generator were evaluated.NRC-DGHW10&FW2.85E-011.0085Failure to recover DG hardware failure or start FW in 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />sThis term represents a failure of a human action to recover DG hardware failure or start FW in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Phase II SAMAs 5 and 8 to install an additional diesel or gas turbine generator were evaluated.

NRC-DG-HW10HRS5.00E-011.005Failure to recover diesel generator hardware failure in 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />sThis term represents a failure of a human action to recover DG hardware failure in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Phase II SAMAs 5 and 8 to install an additional diesel or gas turbine generator were evaluated.NRC-DG-HW1HR9.00E-011.0107Failure to recover diesel generator hardware failure in 1 hourThis term represents a failure of a human action to recover DG hardware failure in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Phase II SAMAs 5 and 8 to install an additional diesel or gas turbine generator were evaluated.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-12NRC-DG-MA1HR9.00E-011.0197Failure to recover diesel generator from maintenance in 1 hourThis term represents a failure of a human action to recover DG hardware failure in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Phase II SAMAs 5 and 8 to install an additional diesel or gas turbine generator were evaluated.

NRC-FO-ADSBOTTLE1.30E-031.0447Failure to connect air bottles to SRV accumulatorsThis term represents a failure of a human action to connect air bottles to the SRV accumulators. Phase II SAMAs 22 and 37 to add larger accumulators and improve SRV pneumatic components were evaluated.NRC-FO-FWS8HR1.10E-021.0107Failure to align FPW for long term injectionThis term represents a failure of a human action to align the firewater system for injection. Phase II SAMAs 24 and 25 for Improved low pressure injection capability were evaluated.NRC-FO-FWSACT5.70E-011.0927Failure to align FPW for long term injectionThis term represents a failure of a human action to align the firewater system for injection. Phase II SAMAs 24 and 25 for Improved low pressure injection capability were evaluated.NRC-OSP-CNT1.21E-021.0052Fail to recover OSP given long term containment failureThis term represents a failure to recover offsite power. Phase II SAMAs 1, 2, 3, 11, 12, 14 and 15 for extending available recovery time by improving DC power were evaluated.NRC-OSP-DLG01.28E-011.0135Fail to recover OSP given 0FTR

failuresThis term represents a failure to recover offsite power. Phase II SAMAs 1, 2, 3, 11, 12, 14 and 15 for extending available recovery time by improving DC power were evaluated.NRC-OSP-DSG06.18E-011.3513Fail to recover OSP given U2

failuresThis term represents a failure to recover offsite power. Phase II SAMAs 1, 2, 3, 11, 12, 14 and 15 for extending available recovery time by improving DC power were evaluated. NRC-OSP-DSG0SSW02.62E-011.0058Fail to recover OSP given U2

FTSThis term represents a failure to recover offsite power. Phase II SAMAs 1, 2, 3, 11, 12, 14 and 15 for extending available recovery time by improving DC power were evaluated.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-13NRC-OSP-DSG11.05E-011.0855Fail to recover OSP given U2

failuresThis term represents a failure to recover offsite power. Phase II SAMAs 1, 2, 3, 11, 12, 14 and 15 for extending available recovery time by improving DC power were evaluated.NRC-OSP-DSG24.53E-021.0126Fail to recover OSP given U2

failuresThis term represents a failure to recover offsite power. Phase II SAMAs 1, 2, 3, 11, 12, 14 and 15 for extending available recovery time by improving DC power were evaluated.NRC-OSP-PSG07.63E-011.0134Fail to recover OSP given SRV LOCA

  • U2
  • No SSW PHV failuresThis term represents a failure to recover offsite power. Phase II SAMAs 1, 2, 3, 11, 12, 14 and 15 for extending available recovery time by improving DC power were evaluated.NRC-SSWPH-VENT3.80E-041.0054Failure to install alternate means of cooling to SSW pump houseThis term represents a failure of a human action to install alternate means of cooling to the SSW pump house. Phase II SAMA 58 for increasing training emphasis and providing control room indication on status of the SSW pump house HVAC was evaluated.NR-PCS-60MN6.00E-011.0402Failure to recover PCS in 60minutesThis term represents a failure of a human action to restore feedwater and manually depressurize. Phase II SAMA 20 for improving high pressure injection capability was evaluated.NRS-ALT-PWR-SUP4.50E-041.0068Failure to align alternate power to 4.16 kV or 6.9 kV busesThis term represents a failure of a human action to align alternate power to 4.16 kV or 6.9 kV buses. Phase II SAMA 6 for improving 4.16kV bus cross-tie ability was evaluated.NRS-DEP-LONG1.20E-051.1703Failure to manually depressurize with ADS/

SRVs (after more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />)This term represents a failure of a human action to manually depressurize with ADS/SRVs after more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Phase II SAMAs 22 and 37 to add larger accumulators and improve SRV pneumatic components were evaluated.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-14NRS-DEP-SHORT3.20E-041.1895Failure to manually depressurize with ADS/

SRVsThis term represents a failure of a human action to manually depressurize with ADS/SRVs. Phase II SAMAs 22 and 37 to add larger accumulators and improve SRV pneumatic components were evaluated.NRS-FO-SSWIA2.20E-041.0104Failure to align SSW B to cool IA compressorsThis term represents a failure of a human action to align SSW B to cool IA compressors. Phase II SAMA 35 for adding IA compressors which do not require cooling was evaluated.NRS-PCS&DEP4.20E-051.037Failure to restore feedwater and manually depressurizeThis term represents a failure of a human action to restore feedwater and manually depressurize. Phase II SAMA 20 for improving high pressure injection capability was evaluated.NRS-PCSL8&DEP1.70E-051.0146Failure to restore feedwater and manually depressurizeThis term represents a failure of a human action to restore feedwater and manually depressurize. Phase II SAMA 20 for improving high pressure injection capability was evaluated.NRS-Y47&FPW2.20E-041.0074Failure of SSW ventilation and align FPWThis term represents a failure of a human action to restore SSW ventilation and align FPW. Phase II SAMA 58 for increased training on restoring SSW ventilation and aligning FPW was evaluated.OSP-LF-EVENTU21.00E+007.1015RCIC failureThis term is a flag. No SAMAs need to be alignedP11.13E-021.0087One stuck-open relief valveThis term represents stuck-open safety relief valves. Phase II SAMA 53 for increased SRV seating reliability was evaluated.P11-PG-XVF021-G7.20E-051.007CST suction manual valve P11-F021 plugsThis term represents a blocked suction for both HPCS and RCIC. Phase II SAMA 20 for adding alternate high pressure injection systems was evaluated.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-15P21.52E-031.0285Two or more stuck-open relief valvesThis term represents stuck-open safety relief valves. Phase II SAMA 53 for increased SRV seating reliability was evaluated.P41-004-A6.43E-031.0096Hardware failure of DG A jacket cooler componentsThis term represents a failure of cooling water to EDGs. Phase II SAMAs 21 and 22 for adding a backup source of diesel cooling were evaluated.P41-054-B6.43E-031.0092Hardware failure of DG B jacket cooler componentsThis term represents a failure of cooling water to EDGs. Phase II SAMAs 9 and 10 for adding a backup source of diesel cooling were evaluated.P41-152-L6.87E-031.0094Hardware failure of RHR heat exchanger coolers TrainAThis term represents a failure of the train A RHR heat exchanger coolers or isolation valves. Phase II SAMA 62 for bypassing the RHR HX SSW isolation valves was evaluated. P41-CC-MVF001A-R6.30E-031.0273Normally closed motor operated valve F001A fails to openThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated. P41-CC-MVF001B-R6.30E-031.0217Normally closed motor driven valve F001B fails to openThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated. P41-CC-MVF005A-R6.30E-031.0273Normally closed motor driven valve F005A fails to openThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated. P41-CC-MVF005B-R6.30E-031.0218Normally closed motor driven valve F005B fails to openThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-16P41-CC-MVF014B-L6.30E-031.0091Motor operated valve F014B-B fails to openThis term represents a failure of the RHR HX SSW isolation valves. Phase II SAMA 62 for bypassing the RHR HX SSW isolation valves was evaluated.

P41-CF-CVDISCH-R1.02E-051.0057Common cause failure of SSW discharge check valvesThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-CF-FNC003S-R3.66E-051.013CCF of 3 or more SSW cooling tower fans to startThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-CF-MVDISCH-R6.99E-051.0407CCF of SSW discharge MOVs FOO5B, FOO5A, &

F011C to openThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.P41-CF-MVF001AB1.85E-041.0117CCF of isolation valves F001A and B to openThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.P41-CF-MVF005AB1.85E-041.0069CCF of discharge MOVs F005A and B to openThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-CF-MVF14AB-L1.85E-041.0056CCF of 2 of 2 SSW RHR HX valves to openThis term represents a failure of the RHR HX SSW isolation valves. Phase II SAMA 62 for bypassing the RHR HX SSW isolation valves was evaluated.

P41-CF-MVF68AB-L1.85E-041.0056CCF of 2 of 2 SSW RHR HX valves to openThis term represents a failure of the RHR HX SSW isolation valves. Phase II SAMA 62 for bypassing the RHR HX SSW isolation valves was evaluated.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-17P41-CF-ST-SUCT-R1.40E-051.0082 of 2 SSW suction strainers CCF to plugThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.P41-FF-MLABST-R1.98E-041.0124Common cause start failures of SSW pumps A & BThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-FF-MLC002C-R9.03E-041.0099Train C pump start failuresThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.P41-FF-MLTCVLV-R6.34E-031.0777SSW Train C common valve hardware failuresThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-FR-MPC002C-R7.20E-041.0063Motor driven pump C002C fails to continue runningThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.P41-LF-FNSSWABC1.00E+001.0207Logic flag-SSW cooling tower fans failThis term is a flag. No SAMAs need to be aligned.P41-MA-SSWA-R2.53E-031.0069SSW Train A unavailable due to maintenanceThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.P41-MA-SSWB-R3.42E-031.0093SSW Train B unavailable due to maintenanceThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-18P41-MA-SSWC-R3.84E-031.0386SSW Train C unavailable due to maintenanceThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.P41-PG-ST-SUCTAR1.39E-041.0089Suct. source failure of motor pumps A & CThis term represents a failure of cooling water to ECCS and PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P53-FO-HECOOLIAS1.00E+001.011Operator fails to align SSW-B to IAS compressor upon loss of TBCWThis term represents a failure of a human action to align SSW-B to the IAS compressor upon loss of TBCW. Phase II SAMA 35 for adding IA compressors which do not require cooling was evaluated.P64-FO-HE-G1.00E+001.1242Operator fails to align firewater system for injectionThis term represents a failure of a human action to align the firewater system for injection. Phase II SAMAs 24 and 25 for improved low pressure injection capability were evaluated.P64-LF-FGSHORT1.00E+001.0471Flag for transient sequences utilizing firewaterThis term is a flag. No SAMAs need to be aligned.P75-CF-3DGR-Z2.16E-041.007CCF of all 3 EDGs to runThis term represents a common cause failure to run of 3 EDGs. Phase II SAMAs 5, 8, 9, 10, and 14 for improvi ng EDG reliability or adding additional onsite power sources were evaluated.P75-CF-3DGS-Z1.53E-051.0054CCF of all 3 EDGs to startThis term represents a common cause failure to start of 3 EDGs. Phase II SAMAs 5, 8, 9, 10, 14, and 16 for improving EDG reliability or adding additional onsite power sources were evaluated.P75-FR-DG-DG11-A4.69E-021.0124DG11 fails to runThis term represents a failure of DG11 to run. Phase II SAMAs 5, 8, 9, 10, 14, and 15 for improving EDG reliability or adding additional onsite power sources were evaluated.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-19P75-FR-DG-DG12-B4.69E-021.016DG12 fails to runThis term represents a failure of DG11 to run. Phase II SAMAs 5, 8, 9, 10, 14 and 15 for improving EDG reliability or adding additional onsite power sources were evaluated.P75-FS-DG-DG11-A6.94E-031.0094DG11 fails to startThis term represents a failure of DG11 to run. Phase II SAMAs 5, 8, 9, 10, 14 and 15 for improving EDG reliability or adding additional onsite power sources were evaluated.P75-FS-DG-DG12-B6.94E-031.0092DG12 fails to startThis term represents a failure of DG11 to run. Phase II SAMAs 5, 8, 9, 10, 14 and 15 for improving EDG reliability or adding additional onsite power sources were evaluated. P75-MA-DGDG11-A1.34E-021.0099DG11 in maintenance outageThis term represents a failure of DG11 to run. Phase II SAMAs 5, 8, 9, 10, 14 and 15 for improving EDG reliability or adding additional onsite power sources were evaluated. P75-MA-DGDG12-B1.19E-021.013DG12 in maintenance outageThis term represents a failure of DG11 to run. Phase II SAMAs 5, 8, 9, 10, 14 and 15 for improving EDG reliability or adding additional onsite power sources were evaluated.P81-FR-DG-DG13-C4.66E-021.0184DG13 fails to runThis term represents a failure DG13 to run. Phase II SAMAs 5, 8, 9, 10, and 14 for improving EDG reliability or adding additional onsite power sources were evaluated.P81-FS-DG-DG13-C5.97E-031.0101DG13 fails to startThis term represents a failure of DG13 to start. Phase II SAMAs 5, 8, 9, 10, 14, and 16 for improving EDG reliability or adding additional onsite power sources were evaluated. P81-MA-DGDG13-C1.18E-021.0152Diesel generator DG13 unavailable due to maintenanceThis term represents maintenance of DG13. Phase II SAMAs 5, 8, 9, 10, and 14 for improving EDG reliability or adding additional onsite power sources were evaluated.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-20R20-CF-CB-BKR2.71E-071.0068CCF of feeder breakers to LCCs feeding the chargers 152-1604 & 1507This term represents a failure of DG11 to run. Phase II SAMAs 5, 8, 9, 10, 14 and 15 for improving EDG reliability or adding additional onsite power sources were evaluated.R20-CF-TR15-169.02E-071.0507CCF of LCC transformers for the 15AA AND 16AB busesThis term represents a failure of the LCC transformers for the 15AA and 16AB buses Phase II SAMAs 5, 8, and 17 to install an additional generator or provide alternate feeds to essential loads from an alternate emergency bus were evaluated.R20-CO-CB-1604-B8.40E-061.0053Feeder breaker 152-1604 fails openThis term represents a failure of the power to LCC 16BB1 and 16BB3. Phase II SAMAs 5, 8, and 17 to install an additional generator or provide alternate feeds to essential loads from an alternate emergency bus were evaluated.

R21-FO-HEESFTRM1.00E+001.0202Operator fails to transfer to alternate transformerThis term represents a failure of a human action to transfer to the alternate transformer. Phase II SAMAs 5, 6, 8, 9, 10, 14, 16, 17, and 18 for enhancing AC system reliability were evaluated.T51-MA-CUB001-C2.00E-031.0188Fan cooler T51B001-C unavailable due to maintenanceThis term represents a failure of the HPCS pump room cooler. Phase II SAMA 29 for adding HPCS HVAC procedures or hardware was evaluated.X31.00E+001.0061X3--depressurization via RCICThis term represents a failure to depressurize with RCIC during a SBO. Phase II SAMAs 1, 2, 3, 11, and 12 for adding or extending battery capacity were evaluated.X77-FF-CFSTART-U4.85E-041.2754X77 common cause start failuresThis term represents a failure of EDG area ventilation. Phase II SAMAs 30, 32, and 33 for adding or enhancing EDG HVAC hardware were evaluated.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-21X77-FF-FSC001A-U3.84E-031.0055DG11 room vent start failuresThis term represents a failure of EDG area ventilation. Phase II SAMAs 30, 32, and 33 for adding or enhancing EDG HVAC hardware were evaluated.X77-FF-FSC001B-U3.84E-031.0053DG12 room vent start failuresThis term represents a failure of EDG area ventilation. Phase II SAMAs 30, 32, and 33 for adding or enhancing EDG HVAC hardware were evaluated.X77-FF-FSC002C-U3.84E-031.0071DIV 3 DG room vent start faultsThis term represents a failure of EDG area ventilation. Phase II SAMAs 30, 32, and 33 for adding or enhancing EDG HVAC hardware were evaluated.Y47-FF-FSC01AA-U6.43E-031.007Y47 Train A start failuresThis term represents a failure of SSW train A pump house ventilation. Phase II SAMA 58 for increasing training emphasis and providing control room indication on status of the SSW pump house HVAC was evaluated.Y47-FO-HEMOD-U1.00E+001.0143Operator fails to provide alternate coolingThis term represents a failure of a human action to provide alternate cooling to the SSW pump house. Phase II SAMA 58 for increasing training emphasis and providing control room indication on status of the SSW pump house HVAC was evaluated.ZLLOCA1.00E+001.036Large LOCA sequenceThis term is a flag. No SAMAs need to be aligned.ZS1LOCA1.00E+001.024Intermediate LOCA sequencesThis term is a flag. No SAMAs need to be aligned.ZS2LOCA1.00E+001.0084Small LOCA sequencesThis term is a flag. No SAMAs need to be aligned.ZSBO1.00E+001.5093SBO sequence (HPCS DG fails)This term is a flag. No SAMAs need to be aligned.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-22ZT1B1.00E+001.0299SBO sequence (HPCS DG success)This term is a flag. No SAMAs need to be aligned.ZTRAN1.00E+002.2786Transient sequence (no SBO)This term is a flag. No SAMAs need to be aligned.Table E.1-2Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-23CDF UncertaintyThe uncertainty associated with core damage frequency (CDF) was estimated and documented in the GGNS Level 1 Model Revision 3 PSA Summary Report [

E.1-5]. The ratio of the 95 th percentile CDF to the mean is about 2.38. An uncertainty factor of 3 was conservatively selected to determine the internal and external benefit with uncertainty described

in Section 4.21.5.4

.E.1.2PSA Model - Level 2 AnalysisE.1.2.1Containment Performance Analysis The GGNS Level 2 PSA model used for the SAMA analysis is the most recent internal events risk model which reflects power uprate conditions [

E.1-4]. The GGNS Level 2 model includes two types of considerations: (1) a deterministic analysis of the physical processes for a spectrum of severe accident progressions, and (2) a probabilistic analysis component in which the likelihood of the various outcomes are assessed. The deterministic analysis examines the response of the containment to the physical processes during a severe accident. This response is performed by*Utilization of the Modular Accident Analysis Program (M AAP) 4.0.6 code to simulate severe accidents that have been identified as dominant contributors to core damage in the Level 1 analysis, and *Reference calculation of several hydrodynamic and heat transfer phenomena that occur during the progression of severe accidents. Examples include debris coolability, pressure spikes due to ex-vessel steam explosions, scoping calculation of direct containment heating, molten debris filling the pedestal sump and flowing over the drywell floor, containment bypass, deflagration and detonation of hydrogen, thrust forces at reactor vessel failure, liner melt-through, and thermal attack of containment penetrations. The Level 2 analysis examined the dominant accident sequences and the resulting plant damage states (PDS) defined in Level 1. The Level 1 analysis involves the assessment of those scenarios that could lead to core damage. A full Level 2 model was developed for GGNS. The Level 2 model consists of containment event trees (CETs) with functional nodes that represent phenomenological events and containment protection system status. The nodes were quantified using subordinate trees and logic rules. A list of the CET functional nodes and descriptions used for the Level 2 analysis is presented in TableE.1-3.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-24The Large Early Release Frequency (LERF) is an indicator of containment performance from the Level 2 results because the magnitude and timing of these releases provide the greatest potential for early health effects to the public. The frequency calculated is approximately 1.05E-7/ry. LERF represents a fraction (~5.1%) of all release end states. TableE.1-4 provides a correlation between the Level 2 RRW risk significant events (severe accident phenomenon, initiating events, component failures, and operator actions) down to 1.005 identified from the GGNS Probabilistic Risk Assessment (PRA) LERF model and the SAMAs evaluated in Section E.2

.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-25Table E.1-3Notation and Definitions for GGNS CET Functional Nodes DescriptionCET Functional NodeSuccess CriteriaParameter Monitored for Success DeterminationReactor Pressure Vessel (RPV) Depressurization (OP)This function questions whether the operator depressurizes the RPV after core damage but before vessel breach. Success of this action would allow low pressure injection, if available, and would minimize the challenge to containment due to a high pressure RPV rupture.The functional success criterion for this node is defined as having the RPV depressurized (i.e., less than 100 psig) until core melt is arrested in-vessel or until the RPV is breached by debris attack.The success of the depressurization function for the RPV is similar to the criterion established in the Level 1 analysis, i.e., prior to core damage. However, there are additional phenomena (i.e., non-condensable gas generation contributing to a high containment pressure that prevents safety relief valve (SRV) operation, and potentially very high containment temperatures which could fail electrical and mechanical components of the SRVs) which can occur during the accident progression beyond core damage and pose further

challenge to the operator's ability to depressurize the RPV.The success criterion is to depressurize the RPV to less than 100 psig via any of the following: *A single SRV open [MAAP case GG10500A_X].

a*Failure of the primary system due to high temperature during core melt progression.

b *A large or medium LOCA.Other alternatives c may be available but are not credited in this analysis. RPV Pressure (< 100 psig)

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-26Core Melt Arrested In-Vessel (RX)In-vessel recovery or arrest of core melt progression addresses the ability of the operating staff to restore adequate core cooling from the time the end state of the Level 1 PRA occurs (e.g., core temperature > 1800°F) until restoration of water injection make-up cannot prevent the breach of the RPV bottom head by debris.Two primary failure modes have been identified for the RPV in the literature: *Local penetration seal failure due to debris heat up and local failure at welds.*Creep rupture failure of the entire bottom head.Preventing the core melt from progressing outside the RPV requires the timely introduction of water onto the debris and intact fuel assemblies. Both timing and system requirements must be defined as part of the success criteria. There are differences in core melt progression models regarding the ability to recover adequate cooling under different circumstances. These vary from no credit for

retention of debris in-vessel after core melting has begun (MAAP 3.0B), to substantial credit for recovery even after debris has accumulated in the bottom head (MAAP 4.0 and MARCH). The best esti mate success criter ia used in this evaluation are based on the time available from the initiation of core degradation until just before substantial core relocation occurs. This typically is on the order of 30-40 minutes. In terms of system requirements, coolant injection is assumed necessary to re-flood the RPV to above 1/3 core height. It is judged, based on deterministic calculations, that this can be accomplished using makeup systems (identified in the Severe Accident Gu idelines (SAGs)) with capability greater than approximately 1000 gpm.

d< 1/2 core relocation calculated by MAAP.Table E.1-3 (Continued)Notation and Definitions for GGNS CET Functional Nodes DescriptionCET Functional NodeSuccess CriteriaParameter Monitored for Success Determination Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-27Igniter Operation (H2) The functional success criterion for this event node is that the igniters operate as designed (1 of 2 divisions). The igniter hardware (1 of 2 divisions), the AC support system, the crew action to initiate, and the H 2 analyzers allowing the initiation are all required for success.

Drywell (DW) Remains Intact (CZ)The functional success criteria for the DW intact node are that the DW retains its pressure capability and that no early DW failure modes compromise the DW integrity. The early DW failures modeled by the CZ node are characterized by phenomenological events (e.g., steam explosions, H 2 deflagration, missile generation, direct containment heating) that are estimated to challenge containment integrity relatively quickly following core melt. Late DW failures, modeled in subsequent nodes, are characterized by extreme pressure and temperature conditions that develop slowly over the course of the accident due to inadequate debris cooling. Note that successful prevention of early DW failure does not necessarily preclude late drywell or containment failure.Therefore, successful prevention of early DW failu re requires the following:*No direct containment heating (direct containment heating is precluded if the RPV is already depressurized).*No ex-vessel steam explosion.

  • No failure of vapor suppression (the suppression pool is not bypassed and no more than 1 drywell to wetwell vacuum breaker fails open).(cont. below)No energetic events and no DW internal pressure

>65psig. No energetic events and no DW differential pressure

>42psid. Table E.1-3 (Continued)Notation and Definitions for GGNS CET Functional Nodes DescriptionCET Functional NodeSuccess CriteriaParameter Monitored for Success Determination Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-28Drywell Remains Intact (CZ) (cont.)*No in-vessel steam explosion (in-vessel steam explosions are precluded if either the RPV is at high pressure, e.g., greater than 100 psig, or the core does not fragment into fine particles before dropping onto the bottom head).*No high pressure spike sufficient to cause DW failure occurs at the time of vessel melt-through (extreme pressure spikes are precluded if the RPV bottom head penetration fails locally or if the RPV remains at low pressure).*No hydrogen deflagration or detonation (if the containment remains steam inert or effective combustible gas igniters operated successfully, then hydrogen detonation or deflagration is guaranteed not to occur).*Containment water pool remains intact.*Upper pool dump operates as needed for those accident scenarios requiring water to cover the top row of horizontal vents.If these failure modes cannot be prevented, large DW failure is assumed to occur. The failure location is assumed to be in the drywell head region and is classified as a large failure.Containment Remains Intact (CX)The functional success criteria for the containment intact node are that the containment retains its pressure capability and that no early containment failure modes compromise the containment integrity. The early containment failures modeled by the CZ node are characterized by phenomenological events (e.g.,

steam explosions, H 2 deflagration, missile generation, direct containment heating) that are estimated to challenge containment integrity relatively quickly following core melt. Late containment failures modeled in subsequent nodes are characterized by extreme pressure and temperature conditions that develop slowly over the course of the accident due to inadequate debris cooling. (cont. below)No energetic containment failure with internal pressure

>65 psig or the containment profile curve.No containment differential pressure > 42 psid.Table E.1-3 (Continued)Notation and Definitions for GGNS CET Functional Nodes DescriptionCET Functional NodeSuccess CriteriaParameter Monitored for Success Determination Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-29Containment Remains Intact (CX) (cont.) Note that successful prevention of early containment failure does not necessarily preclude late containment failure. Therefore, successful prevention of early containment failure requires the following:*No direct containment heating (direct containment heating is precluded if the RPV is already depressurized).*No ex-vessel steam explosion.

  • No failure of vapor suppression (the suppression pool is not bypassed and no more than 1 drywell to wetwell vacuum breaker fails open).*No in-vessel steam explosion (in-vessel steam explosions are precluded if either the RPV is at high pressure, e.g., greater than 100 psig, or the core does not fragment into fine particles before dropping onto the bottom head).*No high pressure spike sufficient to cause containment failure occurs at the time of vessel melt-through (extreme pressure spikes are precluded if the RPV bottom head penetration fails locally or if the RPV remains at low pressure).*No hydrogen deflagration or detonation (if the containment remains steam inert or effective combustible gas igniters operated successfully, then hydrogen detonation or deflagration is guaranteed not to occur).*No continuous RPV blowdown at high pressure via the SRVs or horizontal vents with the suppression pool temperature above 260°F.If these failure modes cannot be prevented, containment failure is assumed to occur. The failure location is assumed to be probabilistically distributed in either the containment airspace above the Aux. Bldg. or the basemat junction with the containment cylinder and is classified as a large failure.Table E.1-3 (Continued)Notation and Definitions for GGNS CET Functional Nodes DescriptionCET Functional NodeSuccess CriteriaParameter Monitored for Success Determination Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-30Containment Isolation (IS)The success of the containment isolation node (IS) is satisfied if the containment penetrations that communicate between the RPV, drywell, or wetwell atmosphere and the secondary containment (or environment) are "closed and isolated." The criteria used to satisfy this requirement of "closed and isolated" is that no line, hatch, or penetration has an opening greater than 2 inches in diameter. This implies that all containment penetrations are adequately sealed and isolated during the entire accident progression until either (1) a safe stable state is reached, or (2) the accident conditions exceed the ultimate capability of containment as determined in the plant specific evaluation.Failure size (< 2 inch dia.)Drywell Isolation (DL)The success of the drywell isolation node (DL) is that the drywell penetrations that allow communication from inside the DW to outside the DW are "closed and isolated." The criteria used to satisfy this requirement of "closed and isolated" is that no line, hatch, or penetration has an opening greater than 2 inches in diameter.Failure size (< 2 inch dia.)Table E.1-3 (Continued)Notation and Definitions for GGNS CET Functional Nodes DescriptionCET Functional NodeSuccess CriteriaParameter Monitored for Success Determination Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-31Debris Cooling (SI)Success at this node requires that water is available (greater than 1000 gpm) to the core debris at the time of vessel failure or shortly thereafter (< 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />).

Continuous water injection either directed into the failed RPV or into the drywell will provide for the following:*Mitigation of high drywell gas temperatures.*Water overburden to scrub fission products resulting from possible core concrete interaction.*Potential for debris coolability.These are considered substantially mitigated if on a best estimate basis a continuous water supply is available to the debris with a flow rate of greater than 1000 gpm.The active mitigation methods that may provide coolant injection to the debris bed include continued make-up to the RPV and containment flooding.These effects would influence the integrity of the DW. Note that inadequate water injection will be modeled for the purposes of consequence evaluation as inducing a drywell failure high in the DW.However, there are some models that indicate that concrete attack and non-condensable gas generation will not be te rminated even if substantial water injection is available to the debris. The temperatures in the drywell will be acceptable, but continued non-condensable gas generation will occur. MAAP sensitivity analyses with minimum heat transfer between debris and water indicate this is not a LERF contributor.Continued concrete attack of the pedestal can result in pedestal failure and consequential failure of the drywell penetrations if the RPV support by the pedestal is compromised.Flow > 1000 gpmTable E.1-3 (Continued)Notation and Definitions for GGNS CET Functional Nodes DescriptionCET Functional NodeSuccess CriteriaParameter Monitored for Success Determination Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-32Containment Flooding Initiated (FC)Success at this node implies that the containment flooding contingency procedure has been initiated by the operating staff and that a system of adequate flow capacity from external sources is available to implement the procedure. In addition to these two requirements, the instrumentation must be available to initiate the flood operation.This node evaluates the possibility that the operator suspends containment flooding because the staff is unable to maintain containment conditions within prescribed limits described in the Emergency Operating Procedures (EOPs) or

SAGs.Containment venting can have varying degrees of releases associated with it depending on the following:*When in the containment flood process containment venting is possible, but not required if RPV is breached.*Whether success of suppression pool cooling and injection is effective in controlling containment pressure.Success at this juncture in the model is defined as the continuation of the flooding evolution with containment cond itions remaining within the limits of the Primary Containment Pressure Limit (PCPL).MAAP calculations indicate that containment flooding through the RPV, containment cooling return, or containment sprays results in a very low radionuclide release [MAAP GG10522].External flow > 1000 gpmVent > 6 inch dia.Table E.1-3 (Continued)Notation and Definitions for GGNS CET Functional Nodes DescriptionCET Functional NodeSuccess CriteriaParameter Monitored for Success Determination Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-33Containment Pressure Control(see node descriptions HR and VC below)Successful containment pressure control is achieved if either of two functional nodes are successfully satisfied:1. Containment heat removal via pool cooling or2. Containment venting Because these have different potential impacts on the radionuclide releases they are treated in separate nodes (see nodes HR and VC below).1. Cont. pressure < 65 psig2. Cont. pressure < 22.4 psig (Venting)Table E.1-3 (Continued)Notation and Definitions for GGNS CET Functional Nodes DescriptionCET Functional NodeSuccess CriteriaParameter Monitored for Success Determination Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-34Containment Heat Removal (HR) Successful containment pressure control is unattainable using suppression pool cooling if either of the following conditions occurs:*No debris cooling (in-vessel or ex-vessel).*Early containment failure.Residual heat removal (RHR) has the capability to remove heat from containment through the RHR heat exchangers.

e This capability requires the following:*A flow path from the suppression pool.*One low pressure coolant injection pump (LPCI) pump.

  • One LPCI pump heat exchanger.
  • SSW to cool the heat exchanger.
  • A return flow path to the suppression pool, the RPV, or the containment spray.*Bypass of the low RPV water level (2/3 core height) interlock if not using RPV return.Failure at this juncture in the sequence implies insufficient containment heat rejection to the environment and continued decay heat generation which could subject the containment to continued pressurization. This condition may eventually cause structural failure, which could subsequently threaten continued successful core coolant injection.Note that RHR success is a moot point if adequate injection to the core or debris has failed. This is because high temperatures from debris radiative heating or high pressure from non-condensable gases will cause drywell failure and containment failure. (MAAP Case GG10506B)Containment pressure < 65 psigTable E.1-3 (Continued)Notation and Definitions for GGNS CET Functional Nodes DescriptionCET Functional NodeSuccess CriteriaParameter Monitored for Success Determination Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-35Containment Venting (VC) The capability to vent the containment is a valuable supplement to the containment pressure control systems. As pressure and temperature increase, there is decreasing confidence in the ability to maintain the integrity of the containment pressure boundary. By instituting a controlled vent of the containment atmosphere, it is possible to maintain long-term containment integrity by providing a viable means of containment pressure control and heat removal. Venting also constitutes a viable mitigative action to minimize the source term released to the environment.Containment venting is successful if it can remove the excess heat and non-condensable gases from the containment and thereby maintain the containment pressure within acceptable limits.Adequate pressure control can be obtained by containment venting if the following conditions are met:*Reactivity control exists.*No "early" containment failure modes occur.
  • Containment flooding does not eliminate the venting pathways.*Vent pathways can be opened and controlled.Based upon deterministic calculations, a containment vent of approximately 6inches in diameter will provide sufficient vent capability to prevent containment failure for sequences involving the loss of containment heat removal or severe accidents.Currently, no vent capability is considered successful for ATWS failure to scram events.Containment pressure < 22.4 psigTable E.1-3 (Continued)Notation and Definitions for GGNS CET Functional Nodes DescriptionCET Functional NodeSuccess CriteriaParameter Monitored for Success Determination Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-36No Suppression Pool Bypass (SP) This node in the CET is used to characterize the magnitude of radionuclides that may escape the containment if wetwell failure or venting occurs. Success means that radionuclides are directed through the suppression pool.

Subsequent headings address specific release paths. Success in preventing suppression pool bypass requires the following: *Vacuum breakers remain closed.*The suppression pool water level remains above the horizontal vents.

  • The drywell does not rupture or fail. Bypass path < 6 inch dia.Wetwell Airspace Breach (WW) (Scrubbed Release)This node appears after the "No Suppression Pool Bypass" node, i.e., drywell intact. This node distinguishes whether the wetwell failure occurred above or below the wetwell water line. Successfully avoiding a large containment failure requires successful containment heat removal.The probabilistic determination of the location of the failure is determined based on the plant specific structural analysis for slow overpressurization events.No WW water release path >2inch dia.Success (Up Branches) containment failure in the dome (Wetwell Airspace.)Containment Spray (CSS)This node distinguishes radionuclide release magnitude based on the availability of the CSS.1 train of CSS operating Enclosure Building/Auxiliary Building Effective (EB)Preservation of the auxiliary building and enclosure building integrity results in a calculated decontamination factor (DF) using MAAP of > 10.DF > 10 (Not currently modeled in MAAP or in the

CET)a.A plant specific assessment of the Grand Gulf response to a high pressure core melt with a single ADS valve opened when the R PV level reaches top of active fuel. This wa s illustrated in MAAP Case GG10500A_X.b.Primary system failure may be induced by very high internal temperatures generated by molten debris in an un-cooled state within the RPV. Such high temperatures coincident with high RPV pressures may lead to localized failures at weak points high within the RPV.c.Opening MSIVs is not credited because this action is not directed by the EOPs when fuel damage has occurred. Table E.1-3 (Continued)Notation and Definitions for GGNS CET Functional Nodes DescriptionCET Functional NodeSuccess CriteriaParameter Monitored for Success Determination Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-37d.The 1000 gpm criterion is an approximation. There is a comparatively large degree of uncertainty surrounding this issue. However, ORNL and GE calculations seem to indicate that an injection rate close to 1000 gpm initiated at thirty minutes may be sufficien

t. The EPRI Technical Basis Report also indicates that this flow rate is adequate. The flow rate is needed to match both the decay heat and the chemical (exothermic) heat generated during postulated core melt progression scenarios. e.Other modes of containment heat removal are not considered effective because of interlocks or procedural restrictions under severe accident conditions (e.g., RWCU, Main Condenser).

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-38Figure E.1-1GGNS Radionuclide Release Category SummaryNote: See Tables E.1-5 and E.1-6 for a definition of the release categories.

H/I 1.23E-08 1%H/L 8.73E-08 4%M/E 3.49E-07 17%Negligible (NCE) 8.73E-07 44%H/E 1.05E-07 5%M/I 1.73E-07 8%M/L 2.71E-07 13%L/E 4.04E-09 0%L/I 3.34E-08 2%LL/I 2.11E-09 0%L/L 1.32E-07 6%LL/E 2.00E-09 0%LL/L 6.83E-09 0%

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-39Figure E.1-2Summary of GGNS Core Damage Accident Sequences Plant Damage StatesNote: Core Damage Accident Sequences Plant Damage State definitions can be seen in TableE.1-7

.IA 38%IBE 33%IBL 3%ID 11%IIA 8%IIIC 5%IVA 0%IVL 1%IIT 1%IIL 0%IIIA 0%IIIB 0%

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-40Table E.1-4Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)Event NameProbabilityRRWEvent DescriptionDispositionB21-LF-FGCTISO1.00E+001.0054Containment isolation signal presentThis term is a flag. No SAMAs need to be aligned.CX2-PH-CZF-NOTSU5.46E-011.024Containment success during severe phenomena (CZ=F, CL II)This term is a split fraction. No SAMAs need to be aligned.CX2-PH-CZS-NOTSU9.82E-011.0173Containment success during severe phenomena (CZ=S, CL II)This term is a split fraction. No SAMAs need to be aligned.CX--PH-CTCOND-F-5.00E-016.317Probability cont. fails given H2 late ignitionThis term represents a failure to control hydrogen or hydrogen ignition. Phase II SAMAs 44 and 45 for reducing the hydrogen detonation potential were evaluated.CX--PH-H2-DEFGF-1.00E+007.7756Hydrogen deflagration occurs globallyThis term represents a failure to control hydrogen or hydrogen ignition. Phase II SAMAs 44 and 45 for reducing the hydrogen detonation potential were evaluated.CX--PH-H2INVENF-1.00E+007.7756Sufficient hydrogen generated to cause overpressureThis term represents a failure to control hydrogen or hydrogen ignition. Phase II SAMAs 44 and 45 for reducing the hydrogen detonation potential were evaluated.CX-PH-LOOP-30MIN8.00E-011.7986AC power not recovered in 30 minThis term represents a failure to recover offsite power. Phase II SAMAs 1, 2, 3, 11, 12 and 15 for extending available recovery time by improving DC power were evaluated.CX--PH-STEAM--F-9.00E-017.7756Containment not inerted by steamThis term represents a failure to control hydrogen or hydrogen ignition. Phase II SAMAs 44 and 45 for reducing the hydrogen detonation potential were evaluated.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-41CZ2-PH-ID-NOTSU6.60E-011.0284Drywell does not fail due to severe phenomena (IGA=F, CLS ID)This term is a split fraction. No SAMAs need to be aligned.CZ4-PH-IGF-NOTSU9.08E-011.008Drywell does not fail due to severe phenomena (IGA=F)This term is a split fraction. No SAMAs need to be aligned.CZ5-PH-IBE-NOTSU9.13E-011.0871Drywell does not fail due to severe

phenomena (CLASS

IBE)This term is a split fraction. No SAMAs need to be aligned.CZ--PH-2-NOTSU9.87E-011.0177Drywell does not fail due to severe phenomena (CLASS II)This term is a split fraction. No SAMAs need to be aligned.CZ--PH-CRDMELTF-1.00E+001.1012Control rods melt prior to fuel rodsThis term represents a possible reactivity excursion due to control rods melting before the fuel rods. Phase II SAMAs 20, 21, 22, and 28 for improving high pressure injection capability were evaluated.CZ--PH-DWFAIL-F-5.00E-015.3793Conditional probability drywell

fails given deflagrationThis term represents a failure to control hydrogen or hydrogen ignition. Phase II SAMAs 44 and 45 for reducing the hydrogen detonation potential were evaluated.Table E.1-4 (Continued)Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-42CZ--PH-FUELRODF-1.00E-021.1012Fuel rod integrity is maintined during the refloodThis term represents timely restoration of emergency core cooling to arrest the core melt progression in-vessel. Phase II SAMAs 20, 21, 22, and 28 for improving high pressure injection capability were evaluated.CZ--PH-SLCLWL-F-1.00E+001.1012Failure to inject SLC with boron for low

water levelThis term represents a failure of a human action to inject SLC with boron for low water level. Phase II SAMAs 20 and 52 for improving high pressure injection and SLC capability were evaluated.E12-FO-HECS-N1.00E+001.0101Operator fails to actuate containment

sprayThis term represents a failure of a human action to actuate containment spray. Phase II SAMAs 46, 47, and 60 for improving containment vent capability were evaluated.E12-FO-HEECCS-G1.00E+001.0058Operator fails to initiate LP ECCSThis term represents a failure of a human action to initiate low pressure ECCS. Phase II SAMAs 20, 21, 22, and 28 for improving high pressure injection capability were evaluated.E12-FO-HESPC-M1.00E+001.0101Operat or fails to manually align for suppression pool

coolingThis term represents a failure of a human action to manually align for suppression pool cooling. Phase II SAMAs 46 and 47 for improving containment vent capability were evaluated.E61-FO-H2-GB-X1.00E+001.0074Failure to obtain grab sample in SAPsThis term represents a failure of a human action obtain grab sample in SAPs. Phase II SAMAs 44 and 45 for installing a passive hydrogen control system were evaluated.E61-FO-IG-L1-X1.00E+001.2301Failure to initiate igniters before

transition to SAPThis term represents a failure of a human action to initiate igniters before transition to SAP. Phase II SAMAs 44 and 45 for installing a passive hydrogen control system were evaluated.Table E.1-4 (Continued)Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-43E61-FO-MSH13-X1.00E+001.2191Operator fails to energize hydrogen

ignitersThis term represents a failure of a human action to energize the hydrogen igniters. Phase II SAMAs 44 and 45 for installing a passive hydrogen control system were evaluated.EV1.00E+001.0808Early declaration of general emergencyThis term is a flag to represent an early declaration of a general emergency. No SAMAs need to be aligned.G-IGNITION5.38E-011.7464Ignition source available at the incorrect timeThis term represents a failure to control hydrogen or hydrogen ignition. Phase II SAMAs 44 and 45 for reducing the hydrogen detonation potential were evaluated.HI--PH-H2IGSBOF-2.50E-011.7986Random hydrogen ignition given no AC powerThis term represents a failure to control hydrogen or hydrogen ignition. Phase II SAMAs 44 and 45 for reducing the hydrogen detonation potential were evaluated.IGA-PH-ID1-NOTSU4.97E-011.0202Igniters successful (CLASS ID)This term is a split fraction. No SAMAs need to be aligned.IGNITERS-FAIL1.00E+001.0828Igniters are operatingThis term is a flag. No SAMAs need to be aligned.IGNITERS-SUC1.00E+001.0356Ingiters are operatingThis term is a flag. No SAMAs need to be aligned.M41-FO-AVVCNT-Q1.00E+001.0058Operator fails to vent containmentThis term represents a failure of a human action to vent containment. Phase II SAMA 46 for a passive containment vent was evaluated.NRC-L2-DEPB&IG3.38E-051.0059Failure to connect ADS bottles and initiate H2 ignitersThis term represents a failure of a human action to emergency depressurize, igniter initiation in level 1, and igniter initiation in level 2. Phase II SAMAs 44 and 45 for installing a passive hydrogen control system were evaluated.Table E.1-4 (Continued)Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-44NRC-OSP-DSG32.87E-021.0153Fail to recover OSP given U2

  • No SSW PHV failuresThis term represents a failure to recover offsite power. Phase II SAMAs 1, 2, 3, 11, 12 and 15 for extending available recovery time by improving DC power were evaluated.NRS-ALTPW&DEP1.00E-061.0052Failure to align alternate power and depressurizeThis term represents a failure of a human action to align alternate power and depressurize. Phase II SAMAs 1, 2, 3, 11, 12, and 15 for extending available recovery time by improving DC power were evaluated.NRS-DHRLT1.00E-071.0058Failure to initiate SPC and containment

sprayThis term represents a failure of a human action to initiate SPC and containment spray. Phase II SAMA 60 for improved containment heat removal were evaluated.NRS-L2-DEP&IG8.32E-061.0716Failure to depressurize and start H2 ignitersThis term represents the operator to fail the following initiation: Emergency depressurization, ignite r initiation in level 1, and igniter initiation in level 2. Phase II SAMAs 44 and 45 for installing a passive hydrogen control system were evaluated.NRS-L2-DEP&IG&FW3.53E-061.0669Failure to depressurize and start H2 igniters and restart FW pumpsThis term represents the operator to fail the following initiation: Emergency depressurization, igniter initiation in level 1, igniter

initiation in level 2, and failure to restart FW. Phase II SAMAs 44 and 45 for installing a passive hydrogen control system were evaluated.NRS-L2-DEP&IG&PCS1.43E-061.0243Failure to depressurize and start H2 igniters and align PCSThis term represents the operator to fail the following initiation: Emergency depressurization, igniter initiation in level 1, igniter initiation in level 2, and align PCS. Phase II SAMAs 44 and 45 for installing a passive hydrogen control system were evaluated.Table E.1-4 (Continued)Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-45OP--AD-ALTRNT-F-1.00E+001.0577Alternate depress. methods not creditedThis is a term to flag not crediting several primary system depressurization schemes. No SAMAs need to be aligned.OP--OP-DEPRESSH-9.68E-011.0544OP fails to depress given OP failed in LVL1 or loss of DCThis term represents a failure of a human action to depressurize given that the operator failed in the level 1 model or a loss of DC. Phase II SAMAs 1, 2, 3, 11, 12, and 15 for extending available recovery time by improving DC power were evaluated.OP--PH-OP1-NOTSU7.11E-011.1331Successful RPV depressurization (Class IA, IE)This term is a split fraction. No SAMAs need to be aligned.OP--PH-OP6-NOTSU9.75E-011.0801Successful RPV depressurization (Class II)This term is a split fraction. No SAMAs need to be aligned.OP--PH-PRESBK-F-8.00E-011.0577Pressure transient does not fail mechanical systemsThis term represents a high pressure vessel breach scenario where mechanical stress failures of the primary system pressure boundary failed to depressurize the RPV. There are no applicable SAMAs for this scenario.OP--PH-SORV---F-5.50E-011.0577SRVs do not fail open during core melt progressionThis term represents a high pressure vessel breach scenario where the SRVs failed to stick open and allow depressurization. There are no applicable SAMAs for this scenario.OP--PH-TEMPBK-F-7.00E-011.0577High prim sys temp does not cause fail of RCS press. boundThis term represents a high pressure vessel breach scenario where the RPV pressure boundary did not rupture due to high internal RPV pressure and temperature. There are no applicable SAMAs for this scenario.Table E.1-4 (Continued)Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-46P41-CF-MV-DGIN-R1.85E-041.0065CCF of DG inlet isol MOVs FO18A-A AND F018B-B to openThis term represents a failure EDG cooling water due to isolation valve failures on the EDG. Phase II SAMAs 5, 8, 9, and 10 for adding an additional generator and increasing the reliability of EDG cooling water were evaluated.P64-PH-RX-EXO-F-1.00E+001.2536FPS (Paths 1-8) inadequate for 1000gpm for Rx nodeThis term is a flag. No SAMAs need to be aligned.P75-CF-DGR-Z1.43E-031.0104CCF of Div 1 & Div 2 (& not Div 3) EDGs to runThis term represents a failure of the emergency AC power. Phase II SAMAs 5, 8, 9, 10, 11, 12 and 14 for improving EDG reliability or adding additional onsite power sources were evaluated.RP--OP-L2-CRODH-1.00E+001.1012Operator restores coolant injec. after ctrl rods are meltedThis term represents a failure of a human action to restore coolant injection after the control rods are melted. Phase II SAMAs 20, 21, 22, and 28 for improving high pressure injection capability were evaluated.RXF1.00E+001.0841Failure of RX (OP=F or Classes IBE, II, IIID, and IV)This is a flag indicating that the RPV is at high pressure with low pressure injection systems not available or viable. No SAMAs need to be aligned.RX--PH-RX2DNOTSU1.09E-011.017Core melt arrested in-vessel (OP=S, Class

ID)This term is a split fraction. No SAMAs need to be aligned.RX--RX-FRECINJH-9.00E-011.2536Operator fails to recover injection before RPV meltThis term represents a failure of a human action to recover injection before the RPV melt. Phase II SAMAs 20, 21, 22, and 28 for improving high pressure injection capability were evaluated.Table E.1-4 (Continued)Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-47Note: Basic events that are correlated in Table E.1-2 are not listed again in Table E.1-4 if they are equivalent basic events.SP--PH-BKFLOW-F-1.00E-011.0352No backflow if SPMU failsThis term represents a suppression pool bypass after a core melt and vessel breach. Phase II SAMA 43 for installing a filtered vent was evaluated.SP--PH-BKIGA-F-1.00E+001.0826No backflow if SPMU failsThis term represents a suppression pool bypass after a core melt and vessel breach. Phase II SAMA 43 for installing a filtered vent was evaluated.SP--VB-SEALS--F-1.00E-021.0178Temperature induced failure of all vacuum breaker sealsThis term represents a suppression pool bypass after a core melt and vessel breach. Phase II SAMA 43 for installing a filtered vent was evaluated.SP--VB-SEALSNWF-5.00E-021.0206Temp induced failure of all vacuum breaker seals (RX=F, SI=F)This term represents a suppression pool bypass after a core melt and vessel breach. Phase II SAMA 43 for installing a filtered vent was evaluated.WW--WW-L2-FAIL--1.00E-021.0196Containment breach below the wtr line (Class I, IIA, IIT, III, IV)This term is a split fraction. No SAMAs need to be aligned.WW--WW-L2-NOT---9.90E-011.2389Containment breach above the wtr line (Class I, IIA, IIT, III, IV)This term is a split fraction. No SAMAs need to be aligned.Table E.1-4 (Continued)Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)Event NameProbabilityRRWEvent DescriptionDisposition Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-48E.1.2.2Radionuclide AnalysisE.1.2.2.1IntroductionA major feature of a Level 2 analysis is the estimation of the source term for every possible outcome of the CET. The CET end points represent the outcomes of possible in-containment accident progression sequences. These end points represent complete severe accident sequences from initiating event to release of radionuclides to the environment. The Level 1 and plant system information is passed through to the CET evaluation in discrete PDS. An atmospheric source term may be associated with each of these CET sequences. Because of the large number of postulated accident scenarios co nsidered, mechanistic calculations (i.e., MAAP calculations) are not performed for every end-state in the CET. Rather, accident sequences produced by the CET are grouped or "binned" into a limited number of release categories, each of which represents all postulated accident scenarios that would produce a similar fission product source term.The criteria used to characterize the release are the estimated magnitude of total release and the timing of the first significant release of radionuclides. The predicted source term associated with each release category, including both the timing and magnitude of the release, is determined using the results of MAAP calculations.E.1.2.2.2Timing of ReleaseTiming completely governs the extent of radioactive decay of short-lived radioisotopes prior to an off-site release and therefore has a first-order influence on immediate health effects. GGNS characterizes the release timing relative to the time at which the release begins, measured from the time of accident initiation. The following three timing categories are used:*Early releases (E) are CET end-states involving containment failure less than 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> from declaration of a general emergency (i.e., prior to effective evacuation), for which minimal offsite protective measures have been observed to be performed in non-nuclear accidents.*Intermediate releases (I) are CET end-states involving containment failure greater than or equal to 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, but less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from declaration of a general emergency, for which much of the offsite nuclear plant protective measures can be assured to be accomplished. *Late releases (L) are CET end-states involving containment failure greater than or equal to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from declaration of a general emergency, for which offsite measures can be assumed to be fully effective.E.1.2.2.3Magnitude of ReleaseSource term results from previous risk studies suggest that categorization of release magnitude based on cesium iodide (CsI) release fractions alone are appropriate [

E.1-7]. The CsI release Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-49fraction indicates the fraction of in-vessel radionuclides escaping to the environment. (Noble gas release levels are non-informative since release of the total core inventory of noble gases is essentially complete given containment failure).The source terms were grouped into five distinct radionuclide release categories or bins according to release magnitude as follows:(1)High (H): A radionuclide release of sufficient magnitude to have the potential to cause early fatalities. This implies a total integrated release of > 10% of the initial core inventory of CsI.(2)Medium (M): A radionuclide release of sufficient magnitude to cause near-term health effects. This implies a total integrated release of between 1% and 10% of the initial core inventory of CsI.(3)Low (L): A radionuclide release with the potential for latent health effects. This implies a total integrated release of between 0.1% and 1% of the initial core inventory of CsI.(4)Low-Low (LL): A radionuclide release with undetectable or minor health effects. This implies a total integrated release of between 0% and 0.1% of the initial core inventory of CsI.(5)Negligible (NCF) - A radionuclide release that is less than or equal to the containment design base leakage.The "total integrated release" as used in the above categories is defined as the integrated release within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after RPV failure. If no RPV failure occurs, then the "total integrated release" is defined as the integrated release within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after accident initiation.E.1.2.2.4Release Category Bin AssignmentsTableE.1-5 summarizes the scheme used to bin sequences with respect to magnitude of release, based on the predicted CsI release fraction and release timing. The combination of release magnitude and timing produce seven distinct release categories for source terms. These are the representative release categories presented in TableE.1-6

.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-50E.1.2.2.5Mapping of Level 1 Results into the Various Release CategoriesPDS provide the interface between the Level 1 and Level 2 analyses (i.e., between core damage accident sequences and fission product release categories). In the PDS analysis, Level 1 results were grouped ("binned") according to plant characteristics that define the status of the reactor, containment, and core cooling systems at the time of core damage. This ensures that systems important to core damage in the Level 1 event trees and the dependencies between containment and other systems are handled consistently in the Level 2 analysis. A PDS therefore represents a grouping of Level 1 sequences that defines a unique set of initial conditions that are likely to yield a similar accident progression through the Level 2 CETs and the attendant challenges to containment integrity.Table E.1-5Release Severity and Timing Classification Scheme SummaryRelease SeverityRelease TimingClassification CategoryCs Iodide % in ReleaseClassification CategoryTime of Initial Release Relative to Time for General Emergency DeclarationHigh (H)Greater than 10Late (L)Greater than 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sMedium or Moderate (M)1 to 10Intermediate (I)4.0 to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sLow (L)0.1 to 1Early (E)Less than 4.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br />sLow-low (LL)Less than 0.1Negligible (NCF)0Table E.1-6GGNS Release CategoriesTime of ReleaseMagnitude of ReleaseHMLLLEH/EM/EL/ELL/EIH/IM/IL/ILL/ILH/LM/LL/LLL/L Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-51From the perspective of the Level 2 assessment, PDS binning entails the transfer of specific information from the Level 1 to the Level 2 analyses.*Equipment failures in Level 1. Equipment failures in support systems, accident prevention systems, and mitigation systems that have been noted in the Level 1 analysis are carried into the Level 2 analysis. In this latter analysis, the repair or recovery of failed equipment is not allowed unless an explicit evaluation, including a consideration of adverse environments where appropriate, has been performed as part of the Level 2 analysis.*RPV status. The RPV pressure condition is explicitly transferred from the Level 1 analysis to the CET.*Containment status. The containment status is explicitly transferred from the Level 1 analysis to the CET. This includes recognition of whether the containment is bypassed or is intact at the onset of core damage.*Differences in accident sequence timing are transferred with the Level 1 sequences. Timing affects such sequences as: SBO, internal flooding, and containment bypass (ISLOCA).This transfer of information allows timing to be properly assessed in the Level 2 analysis.Based on the above criteria, the Level 1 results were binned into PDS. These PDS define important combinations of system states that can result in distinctly different accident progression pathways and therefore, different containment failure and source term characteristics. TableE.1-7 provides a description of the GGNS PDS that are used to summarize the Level 1 results.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-52Table E.1-7Summary of GGNS Core Damage Accident Sequences Plant Damage StatesAccident Class DesignatorSubclassDefinitionCAFTAModel(per Rx Yr)Class IAAccident sequences involving loss of inventory makeup in which the reactor pressure remains high.1.12E-06BAccident sequences involving a station blackout and loss of coolant inventory makeup. (Class IBE is defined as "Early" Station Blackout events with core damage at less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Class IBL is defined as "Late" Station Blackout events with core damage at greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.)IBE = 9.71E-07IBL = 8.20E-08CAccident sequences involving a loss of coolant inventory induced by an ATWS sequence with containment intact.< 1E-12DAccident sequences involving a loss of coolant inventory makeup in which reactor pressure has been successfully reduced to 200 psi.3.17E-07EAccident sequences involving loss of invent ory makeup in which the reactor pressure remains high and DC power is unavailable. (Grouped with Class IA.)(Grouped with Class IA)Class IIAAccident sequences involving a loss of containment heat removal with the RPV initially intact; core damage; core damage induced post containment failure. 2.44E-07LAccident sequences involving a loss of containment heat removal with the RPV breached but no initial core damage; core damage induced post containment failure.9.84E-10TAccident sequences involving a loss of containment heat removal with the RPV initially intact; core damage induced post high containment pressure.1.37E-08VClass IIA and IIT except that the vent operates as designed; loss of makeup occurs at some time following vent initiation. Suppression pool saturated but intact.< 1E-12 Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-53Note: The total CDF is not the same as the baseline CDF in Table E.1-1 due to non-minimal cutsets created when quantifying at the sequence level.Class III (LOCA)AAccident sequences leading to core damage conditions initiated by vessel rupture where the containment integrity is not breached in the initial time phase of the accident. 1.00E-08BAccident sequences initiated or resulting in small or medium LOCAs for which the reactor cannot be depressurized prior to core damage occurring. 6.39E-11CAccident sequences initiated or resulting in medium or large LOCAs for which the reactor is at low pressure and no effective injection is available. 1.60E-07DAccident sequences which are initiated by a LOCA or RPV failure and for which the vapor suppression system is inadequate, challenging the containment integrity with subsequent failure of makeup systems. < 1E-12 Class IV (ATWS)AAccident sequences involving failure of adequate shutdown reactivity with the RPV initially intact; core damage induced post containment failure. 4.06E-09LAccident sequences involving a failure of adequate shutdown reactivity with the RPV initially breached (e.g., LOCA or stuck-open relief valve (SORV)); core damage induced post containment failure. (Grouped with Class IVA)Class V---Unisolated LOCA outside containment.4.91E-10Total CDF2.92E-06Table E.1-7 (Continued)Summary of GGNS Core Damage Accident Sequences Plant Damage StatesAccident Class DesignatorSubclassDefinitionCAFTAModel(per Rx Yr)

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-54E.1.2.2.6Process Used to Group the Source TermsThe approach used to evaluate radionuclide releases and develop release categories is similar to that applied in the NUREG-1150 [

E.1-8] analysis. The objectives were to establish the timing of the first significant release of radionuclides and estimate the magnitude of the total release.The GGNS Level 3 analysis requires, as an input, the frequency, type, timing and amount of fission products released to the environment during the core damage accidents postulated by the GGNS Level 2 PRA analyses. In order to simplify the large number of potential release scenarios, a representative set of release fractions was chosen for each containment event tree end state along with an end state frequency.The PDS designators listed in TableE.1-7 represent the core damage end state categories from the Level 1 analysis that are grouped together as entry conditions for the Level 2 analysis. The Level 2 accident progression for each of the PDS is evaluated using a CET to determine the appropriate release category for each Level 2 sequence. Note, however, that since not all the Level 2 sequences associated with each Level 1 plant damage state may be assigned to the same release category, there is no direct link between a specific Level 1 core damage PDS and Level 2 release category. Rather, the sum of the Level 2 end state frequencies assigned to each release category determines the overall frequency of that release category.Appendix D of the GGNS Level 2 PSA Analysis [

E.1-4] describes which GG NS specific MAAP analyses are representative of each CET end state. It also bins each CET sequence into one of the release categories depicted in TableE.1-6

.For each CET sequence, a value for each of the release-to-environment mass fractions was obtained from the representative MAAP calculation. These mass fractions were then weighed according to the contribution of that sequence to the sum of the sequences in the end state bin. The final mass fraction representing the end state bin was the sum of these individual weighed mass fractions for each species.To evaluate the Level 2 model results in a manner that provided the above information, each Level 2 CET sequence was linked to its respective CET end state (H/E, H/I, H/L, etc.). The release fraction and timing data for all sequences associated with a particular CET end state were weighted according to the sequence weight for that end state and summed to obtain a representative release fraction and release timing for that end state.Based on the above binning methodology, the salient Level 2 results are summarized in TableE.1-8. Table E.1-8 summarizes the results of the CET quantification and identifies the total annual release frequency for each Level 2 release category.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-55Nomenclature:Timing (time between General Emergency Declaration and initial release):Late (L) - Greater than 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sIntermediate (I)- 4.0 to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sEarly (E)- Less than 4.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br />sMagnitude:Negligible (NCF)- Much less than 0.1% CsI release fractionLow-Low (LL)- Less than 0.1% CsI release fractionLow (L) - 0.1% to 1% CsI release fractionMedium (M) - 1% to 10% CsI release fraction High (H) - Greater than 10% CsI release fractionTable E.1-8Summary of Containment Event Tree Quantification Release Category(Magnitude/Timing)Release Frequency(Per ry)H/E1.05E-07H/I1.23E-08H/L8.73E-08M/E3.49E-07M/I1.73E-07M/L2.71E-07L/E4.04E-09L/I3.34E-08L/L1.32E-07LL/E2.00E-09LL/I2.11E-09LL/L6.83E-09Negligible (NCF)8.73E-07CDF2.05E-06 Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-56E.1.2.2.7Consequence Analysis Source TermsInput to the Level 3 GGNS model from the Level 2 model is a combination of radionuclide release fractions, timing of radionuclide releases, and frequencies at which the releases occur. This combination of information is used in conjunction with GGNS site characteristics in the Level 3 model to evaluate the off-site consequences of a core damage event. Source terms were developed for the release categories identified in TableE.1-6. TableE.1-9 provides a summary of the Level 2 results that were used as Level 3 input for the GGNS SAMA analysis (the baseline analysis case). Consequences corresponding to each of the release categories are developed in the GGNS Level 3 model, which is discussed in Section E.1.5

.E.1.2.2.8Release Magnitude CalculationsThe MAAP computer code is used to assign both the radionuclide release magnitude and timing based on the accident progression characterization. Specifically , MAAP provides the following information:*Containment pressure and temperature (time of containment failure is determined by comparing these values with the nominal containment capability).*Radionuclide release timing and magnitude for a large number of radioisotopes.*Release fractions for twelve radionuclide species.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-57Table E.1-9GGNS Release Category Source TermsSheet 1 of 2Release Mode (CET End State)Frequency (/year)Warning Time (sec)Elevation (m)Release Start (sec)Release Duration (sec)Release Energy (W)

H/E1.05E-079663214972577033.0E+07H/I1.23E-081232377072214934.3E+05 H/L8.73E-08992321120961111049.1E+04 M/E3.49E-0795732596641995362.9E+06 M/I1.73E-071232908661683344.3E+05 M/L2.71E-07992321169441422569.1E+04 L/E4.04E-0978732127947332.4E+06L/I3.34E-0812643219972572034.3E+05 L/L1.32E-07966321070781521228.3E+05 LL/E2.00E-0912663219962572042.9E+06LL/I2.11E-0912653219962572044.3E+05 LL/L6.83E-09126632186290729109.1E+04 Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-58Table E.1-9GGNS Release Category Source TermsSheet 2 of 2Release Mode (CET End State)Release FractionNGICsTeSrRuLaCeBa H/E1.0E+001.4E-026.1E-035.5E-024.0E-046.5E-054.7E-055.0E-041.9E-04H/I1.0E+002.2E-017.6E-021.3E-019.8E-061.1E-055.4E-079.6E-061.0E-05 H/L1.0E+001.7E-014.7E-021.5E-015.9E-065.7E-076.7E-077.3E-063.6E-06 M/E8.8E-011.8E-015.2E-021.1E-011.5E-039.7E-041.4E-041.7E-031.2E-03 M/I1.0E+003.6E-021.5E-011.1E-013.6E-061.1E-052.6E-074.5E-069.9E-06 M/L1.0E+008.4E-025.0E-024.9E-022.2E-076.2E-071.6E-082.0E-071.3E-06 L/E9.1E-012.1E-032.1E-032.1E-031.2E-053.9E-042.6E-071.5E-066.4E-05L/I1.0E+008.3E-022.5E-026.9E-021.7E-044.2E-054.4E-061.3E-049.2E-05 L/L1.0E+007.2E-034.5E-034.3E-024.4E-061.4E-065.0E-076.1E-064.8E-06 LL/E2.1E-025.3E-065.4E-071.9E-062.6E-092.3E-072.0E-101.1E-096.5E-08LL/I1.9E-021.7E-062.6E-073.0E-061.6E-092.2E-071.2E-106.6E-104.7E-08 LL/L9.6E-011.0E-021.7E-021.0E-021.5E-061.8E-061.3E-071.2E-061.0E-06 Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-59E.1.3IPEEE AnalysisE.1.3.1Seismic AnalysisThe seismic portion of the IPEEE was completed in December 1994 and documented in GGNS94-0053 [

E.1-9] following the guidance of NUREG-1407 [E.1-10] and EPRI NP-6041-SL

[E.1-11]. The SMA approach is a deterministic and conservative evaluation that does not calculate risk on a probabilistic basis. Therefore, its results should not be compared directly with the best-estimate internal events results.

The conclusions of the GGNS IPEEE seismic margin analysis are as follows:*Walkdowns resulted in no outliers that are operability issues at the plant.*No unique decay heat removal vulnerabilities to seismic events were found. *Seismic-induced flooding and fires do not pose major risks.*No unique seismic-induced containment failure mechanisms were identified.A number of plant improvements were identified and resolved as a result of the report. The list can be found in Appendix B of GGNS94-0053 Seismic Margins IPEEE [

E.1-9].E.1.3.2Fire Analysis

The GGNS internal fire risk model was performed in the mid-1990's as part of the IPEEE for GGNS. The GGNS fire analysis was performed using EPRI's Fire PRA Implementation Guide

[E.1-12].TableE.1-10 presents the results of current GGNS IPEEE fire analysis.Generic conservatisms in the IPEEE fire analysis methods mentioned in NEI 05-01 [

E.1-1], "Severe Accident Mitigation Alternatives (SAMA) Analysis Guidance Document," that are applicable to the GGNS fire analysis include the following.*The frequency and severity of fires were generally conservatively overestimated. A revised NRC fire events database indicates a trend toward lower frequency and less severe fires. This trend reflects improved housekeeping, reduction in transient fire hazards, and other improved fire protection steps at utilities.*There is little industry exper ience with crew actions following fires. This led to conservative characterization of crew actions in the IPEEE fire analysis. Because CDF is strongly correlated with crew actions, this conservatism has a profound effect on fire results.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-60*The peer review process for fire analyses was less well developed than for internal events PSAs. For example, no industry proc ess, such as NEI 00-02, existed for the structured peer review of a fire PSA.Plant-specific conservative as sumptions in the GGNS IPEEE fire analysis include the following.*Certain specified components whose locations were not determined were assumed failed by any fire.*Plant trip initiators were assumed to occur in each fire area.*The damaging effects of a fire were assumed to affect all components in a compartment unless detailed fire modeling was done to demonstrate otherwise.*No credit was given to human detection except when a continuous fire watch is required.*Suppression prior to loss of a cabinet's function was not credited. This assumption was particularly important to the control room.*The loss of a control room cabinet containing divisional equipment was assumed to affect the entire division.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-61Table E.1-10GGNS Fire IPEEE Results Fire CompartmentCompartment DescriptionTotal Compartment CDF (/rx-yr)

Screened (1)CA101Auxiliary Building Corridors. 93'-0" Elevation 5.74E-07NCA102RB Fire Zones 1A102, 1A202, 1A303, 1A4421.01E-08BCA103RB Fire Zones 1A103, 1A2034.85E-08BCA104RB Fire Zones 1A104, 1A2046.76E-08BCA105RB Fire Zones 1A105, 1A2057.19E-07B CA106RB Fire Zones 1A106, 1A206, 1A307, 1A4418.56E-09BCA107 RB Fire Zones 1A1072.06E-08ACA108RB Fire Zones 1A1082.06E-08A CA109 RB Fire Zones 1A1096.31E-07BCA111RB Fire Zones 1A111, 1A1271.06E-08ACA115RB Fire Zones 1A115, 1A116, 1A118, 1A119, 1A2201.34E-07CCA124RB Fire Zones 1A1247.35E-09ACA125RB Fire Zones 1A1257.35E-09ACA130RB Fire Zones 1A130, 1A1317.35E-09A CA132RB Fire Zones 1A132, 1A224, 1A226, 1A305, 1A439, 1A4408.83E-09BCA201Auxiliary Building Corridors. 119'-0" Elevation 6.38E-07NCA207Switchgear Room 1A2073.47E-07CCA208Switchgear Room 1A2088.14E-07CCA209RB Fire Zones 1A2091.41E-08A CA210RB Fire Zones 1A2101.41E-08ACA219Switchgear Room 1A2194.09E-07CCA221Switchgear Room 1A2214.57E-07C CA225RB Fire Zones 1A2257.35E-09ACA301Auxiliary Building Corridors. 139'-0" Elevation A422, 1A3246.70E-07NCA304RB Fire Zones 1A3047.36E-09BCA306RB Fire Zones 1A3067.36E-09BCA308RB Fire Zones 1A3082.62E-08BCA309RB Fire Zones 1A3093.57E-07B Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-62CA318RB Fire Zones 1A3182.41E-07BCA319RB Fire Zones 1A3198.82E-09A CA320RB Fire Zones 1A3203.09E-07BCA323RB Fire Zones 1A3238.82E-09ACA325RB Fire Zones 1A3257.35E-09A CA326RB Fire Zones 1A3261.09E-08ACA401RB Fire Zones 1A401, 1A403, 1A417, 1A420, 1A424, 1A427, 1A428, 1A4341.94E-07CCA402RB Fire Zones 1A4027.35E-09ACA404RB Fire Zones 1A4047.35E-09ACA405RB Fire Zones 1A4052.94E-08ACA406RB Fire Zones 1A4061.12E-08A CA407Switchgear Room 1A4075.00E-08BCA410Switchgear Room 1A4105.00E-08BCA429RB Fire Zones 1A4291.06E-08A CA430RB Fire Zones 1A4302.73E-08ACA431RB Fire Zones 1A431, 1A437, 1A438, 1A444, 1A525, 1A528, 1A532, 1A602, 1A603, 1A604, 1A606, 1A6071.32E-07ACA432RB Fire Zones 1A4322.38E-08ACA433RB Fire Zones 1A4331.41E-08ACA436RB Fire Zones 1A4367.35E-09A CA506RB Fire Zones 1A506, 1A508, 1A6057.35E-09ACA519RB Fire Zones 1A519, 1A523, 1A524, 1A527, 1A5312.26E-08ACA529RB Fire Zones 1A5292.73E-08BCA530RB Fire Zones 1A5302.06E-08ACA533RB Fire Zones 1A5337.35E-09ACA534RB Fire Zones 1A5347.35E-09A CA536RB Fire Zones 1A5367.35E-09ACA537RB Fire Zones 1A5377.35E-09ATable E.1-10 (Continued)GGNS Fire IPEEE Results Fire CompartmentCompartment DescriptionTotal Compartment CDF (/rx-yr)

Screened (1)

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-63CA539RB Fire Zones 1A5393.45E-07CCC101RB Fire Zones OC101, OC103, OC115, OC117, OC2175.67E-08CCC104Hot Machine Shop 2.42E-07NCC125RB Fire Zones OC1251.18E-08ACC126RB Fire Zones OC1262.06E-08A CC128RB Fire Zones OC1281.38E-08BCC202Division 1 Switchgear Room 9.37E-07NCC203RB Fire Zones OC2032.94E-07C CC204RB Fire Zones OC2041.35E-08ACC205RB Fire Zones OC2057.35E-09ACC205ARB Fire Zones OC205A7.35E-09A CC206RB Fire Zones OC2067.35E-09ACC207Battery Room OC2078.84E-07CCC208RB Fire Zones OC2088.75E-08B CC208ARB Fire Zones OC208A3.75E-08BCC209Battery Room OC2094.63E-07BCC210Division 3 (HPCS) Switchgear Room 6.08E-07NCC211Battery Room OC2112.94E-07CCC212RB Fire Zones OC2127.35E-09A CC213RB Fire Zones OC2133.86E-08BCC214RB Fire Zones OC2144.24E-07CCC215Division 2 Switchgear Room 4.06E-07N CC216RB Fire Zones OC2167.36E-09BCC218RB Fire Zones OC2187.35E-09ACC219RB Fire Zones OC2197.35E-09A CC301RB Fire Zones OC3017.35E-09ACC302HVAC Equipment Room2.10E-07NCC303RB Fire Zones OC3034.42E-08B CC304RB Fire Zones OC304, OC412, OC6124.46E-08BTable E.1-10 (Continued)GGNS Fire IPEEE Results Fire CompartmentCompartment DescriptionTotal Compartment CDF (/rx-yr)

Screened (1)

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-64CC305RB Fire Zones OC3057.35E-09ACC306RB Fire Zones OC306, OC409, OC610, OC7091.03E-08ACC307RB Fire Zones OC3071.94E-07CCC308RB Fire Zones OC3087.36E-09BCC309RB Fire Zones OC3097.35E-09A CC401RB Fire Zones OC4017.35E-09ACC402Cable Spreading Room 2.82E-07NCC402ARB Fire Zones OC402A, OC512B7.35E-09A CC403RB Fire Zones OC4031.04E-07CCC404RB Fire Zones OC4047.35E-09ACC405RB Fire Zones OC4057.35E-09A CC405ARB Fire Zones OC405A, OC507A7.35E-09ACC406RB Fire Zones OC4067.35E-09ACC406ARB Fire Zones OC406A, OC518A, OC613A7.35E-09A CC407RB Fire Zones OC4071.35E-07ACC408RB Fire Zones OC4081.05E-07BCC409ARB Fire Zones OC409A, OC512, OC608B7.35E-09ACC410Battery Room OC4101.91E-08ACC411RB Fire Zones OC4117.35E-09A CC412ARB Fire Zones OC412A, OC507C, OC603B7.35E-09ACC501RB Fire Zones OC5017.35E-09ACC502Control Room 3.85E-06N CC507RB Fire Zones OC5077.35E-09ACC509RB Fire Zones OC509, OC511, OC5127.35E-09ACC510RB Fire Zones OC5107.35E-09A CC513RB Fire Zones OC5137.35E-09ACC514RB Fire Zones OC5147.35E-09ACC515RB Fire Zones OC5157.35E-09A CC518RB Fire Zones OC518, OC6117.35E-09ATable E.1-10 (Continued)GGNS Fire IPEEE Results Fire CompartmentCompartment DescriptionTotal Compartment CDF (/rx-yr)

Screened (1)

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-65CC601RB Fire Zones OC601, OC6024.58E-08BCC603RB Fire Zones OC6031.12E-08A CC604RB Fire Zones OC6049.70E-09ACC606RB Fire Zones OC6067.35E-09ACC608RB Fire Zones OC6081.03E-08A CC609RB Fire Zones OC6097.35E-09ACC613RB Fire Zones OC6137.35E-09ACC614RB Fire Zones OC6149.11E-09A CC615RB Fire Zones OC6157.35E-09ACC616RB Fire Zones OC6167.35E-09ACC617RB Fire Zones OC6177.35E-09A CC618RB Fire Zones OC6181.05E-07BCC619RB Fire Zones OC6197.35E-09ACC701RB Fire Zones OC7017.35E-09A CC702Cable Spreading Room OC7025.18E-07CCC703RB Fire Zones OC7034.72E-07CCC704RB Fire Zones OC7042.12E-08A CC705RB Fire Zones OC7058.82E-09ACC706RB Fire Zones OC7061.05E-07BCC707RB Fire Zones OC7071.09E-07ACC708RB Fire Zones OC708, OC7105.59E-08ACC708ARB Fire Zones OC708A7.35E-09A CC711RB Fire Zones OC7117.35E-09ACC712RB Fire Zones OC7127.35E-09ACC713RB Fire Zones OC7137.35E-09A CD301RB Fire Zones 1D3014.34E-08BCD306Division 3 (HPCS) Diesel Generator Room 1.72E-07NCD308Diesel Generator Room ID3083.05E-07C CD310Diesel Generator Room ID3103.48E-07CCM100RB Fire Zones BASIN NO. 12.80E-08BTable E.1-10 (Continued)GGNS Fire IPEEE Results Fire CompartmentCompartment DescriptionTotal Compartment CDF (/rx-yr)

Screened (1)

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-66CM110RB Fire Zones IM1102.56E-07CCM112RB Fire Zones IM1127.90E-07B CM200RB Fire Zones BASIN NO. 22.90E-08BCM210RB Fire Zones 2M1107.31E-07BCM212RB Fire Zones 2M1124.02E-08B CT100 Turbine Building Floor, 93'-0" Elevation 3.24E-07NCT200Turbine Building Floor, 113'-0" Elevation 7.10E-09NCT212Battery Room 1T2121.91E-08A CT219Switchgear Room 1T2199.23E-07BCT300133'-0" Elevation, Turbine Bldg.5.19E-07CCT312Battery Room 1T3121.91E-08A CT323Switchgear Room 1T3238.84E-07BCT400166'-0" Elevation, Turbine Bldg. + 1T502, 1T5031.63E-07CCT405Battery Room 1T4051.91E-08ACT406Battery Room 1T4061.91E-08ACM101OM101 (Circ. Water Pumphouse) 6.48E-08BCM102OM102 (Mtr. Driver Fire Pump Room) 1.14E-07BCM115OM115 (all Water Treatment Bldg.) 2.42E-07CCRADRadwaste Bldg.3.29E-07C CTR11Transformers BOP11A, BOP11B1.94E-07CCTR12Transformers BOP12A, BOP12B1.65E-07BCTR14Transformers BOP14, BOP244.70E-08A CTRMAINTransformers Main 1A, 1B, 1C, 1D8.53E-08ACDUC1Division 1 duct bank to SSW Cooling Tower3.15E-08BCDUC2Division 2 duct bank to SSW Cooling Tower2.47E-07C CDUC3Division 3 duct bank to SSW Cooling Tower2.52E-07BYARDBalance of Yard Area 7.71E-07CTotal2.74E-05

Reference:

E.1-15Table E.1-10 (Continued)GGNS Fire IPEEE Results Fire CompartmentCompartment DescriptionTotal Compartment CDF (/rx-yr)

Screened (1)

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-67E.1.3.3Other External HazardsThe GGNS IPEEE submittal, in addition to the internal fires and seismic events, examined a number of other external hazards:*High winds and tornadoes.*External flooding.*Ice, hazardous chemical, transportation, and nearby facility incidents.

The GGNS Individual Plant Examination of External Events (IPEEE) concluded that for high winds, floods, and other external events, GGNS meets the applicable Nuclear Regulatory Commission (NRC) requirements and therefore has an acceptably low risk with respect to these hazards. As these events are not dominant contributors to external event risk and quantitative analysis of these events is not practical, they are considered negligible.1.Screening Criteria in Table E.1-10:AScreened based on no safe shutdown or PRA equipment.

BScreened assuming all equipment in compartment is failed.CScreened with credit for detailed recovery.NNot screened, more detailed analysis performed.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-68E.1.4PSA Model Revisions and Peer Review SummaryThe summary of the GGNS PSA models CDF and LERF is presented in the table below.E.1.4.1Major Differences between the 1997(R1) PSA Model and the IPE ModelThe GGNS IPE model was originally made in 1992 [

E.1-13]. The NRC provided a Safety Evaluation of the IPE in March 1996 [

E.1-16]. It was then updated in 1997 and was renamed the GGNS. A summary of this update is documented in GGNS Engineering Report No. GGNS-97-0014 [E.1-14]. The changes lowered the CDF to 5.46E-06/rx-yr from 1.72E-05/rx-yr. Changes to the model are summarized below:*Incorporation of updated plant specific data for system maintenance and testing unavailability.*Incorporation of updated plant specific data for initiating event frequencies.*Incorporation of updated plant specific data for certain important components (i.e., diesel generators, HPCS and reactor core isolation cooling (RCIC) pumps).*Various modeling changes to system models to correct minor modeling errors and incorporate modifications since the original IPE.E.1.4.2Major Differences between the 2002 (R2) PSA Model and the 1997(R1) PSA ModelThe next update of the PSA model was identified as the GGNS Level 1 PSA, Revision 2. This update included plant changes through refueling outage 11, addition of an ISLOCA initiator, and operating data through 12/31/2000. It is documented in GGNS calculation XC-N1111-01007

[E.1-17]. The changes lowered the CDF to 4.27E-06/rx-yr from 5.46E-06/rx-yr. The LERF calculation was also updated and the results changed to 2.04E-07/rx-yr. Summary of Major PSA ModelsPSA ModelCDF (/rx-yr)LERF (/rx-yr)1992 (IPE)1.72E-055.17E-071997 (R1)5.46E-06Not Updated2002 (R2)4.27E-062.04E-07 2010 (R3)2.69E-061.44E-072010 (EPU)2.91E-061.48E-07 Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-69E.1.4.3Major Differences between the 2010 (R3) PSA Model and the 2002 (R2) PSA ModelThe update of the Revision 2 Model is designated as the GGNS Level 1, Revision 3 Model [

E.1-5]. The following list describes the most significant changes from the 2002 (R2) model.*Updated plant specific data (through 8-2006). *Updated plant specific (through 8-2006) and generic initiator frequencies. *New initiators: Loss of service transformer. Reactor Vessel Rupture. Loss of CRD. Break (LOCA) Outside of Containment. *Major changes to LOSP modeling: Added loss of preferred offsite power initiator. Added consequential loss of offsite power event as a result of transient initiator. Added consequential loss of offsite power event as a result of LOCA initiator. New industry data used for LOSP recovery analysis. *Separated loss of PCS initiator into Closure of MSIVs initiator and Loss of PCS due to other causes initiator. *Updated ISLOCA analysis. *Updated common cause analysis.

  • Updated human reliability analysis. *Included modeling for loss of ECCS pumps due to containment failure. *Revised instrument air system modeling to incorporate new Plant Air compressors.*Revised modeling of CRD-less credit for CRD. *Added more detailed modeling for failure to scram. *Added more detail to power conversion model.The calculation PRA-GG-01-001 [

E.1-5] summarizes changes incorporated in the Revision 3 model, the overall core damage frequency results, and other additional information from the Revision 3 version of the model. These changes lowered the CDF to 2.69E-06/rx-yr from 4.27E-06/rx-yr. The LERF contribution from this model is 1.44E-07/rx-yr.

E.1.4.4Major Differences between the 2010 (EPU) PSA Model and the 2010 (R3) PSA ModelThe 2010 (R3) PSA model is based on the current licensed thermal power (CLTP) level of 3898 MWt. The 2010 EPU model uses a 13 percent increase (i.e., extended power uprate) of the CLTP to 4408 MWt.The Grand Gulf PRA was examined to assess the impact of the following EPU changes on the PRA elements:*Power level change Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-70*Hardware changes*Procedural changes*Operational changesThe results of the PRA evaluation are the following:*Detailed thermal hydraulic analyses of the plant response using the EPU configuration indicate reductions in the operator action "allowable" times for some actions.*The reduced operator action "allowable" times resulted in increases in the assessed human error probabilities for some actions in the PRA model.*Only small risk increases were identified for the changes associated with the EPU. These involved (1) reduced times available for effective operator actions and (2) minor changes in some functional success criteria in the PRA (negligible impact on results).*The risk impact due to the implementation of the EPU is low and acceptable without the requirement for special compensatory measures. The risk impact is in the "very low" category (i.e., Region III) of the Regulatory Guide 1.174 guidelines for CDF and for LERF.The EPU is estimated to increase the Grand Gulf internal events PRA CDF to 2.91E-6/rx-yr, an increase of ~8.6%. In addition a full level 2 model was created which reflects EPU conditions

[E.1-4]. In this model, LERF increased to 1.48E-07/rx-yr, an increase of ~3%.The following table shows the changes in contribution to CDF per initiator group for each model revision.Contribution to CDF Changes in PRA ModelsContributing Initiator GroupR1R2R3R3 EPULOSP42.5%38.6%38.7%39.5%Loss of Feedwater (FW)4.3%21.2%8.6%8.1%PCS Avail Trans5.9%16.1%20.8%20.5%Loss of PCS7.7%12.1%12.9%12.4%Special(1)30.6%11.3%7.9%7.8%LOCA7.9%0.4%4.0%3.8%SORV0.1%0.1%0.3%0.3%

ISLOCAnot modeled0.1%0.0%0.0%Flood1.0%0.1%0.1%0.0%

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-71E.1.4.5PSA Model Peer ReviewThe 1997 (Rev. 1) Level 1 and LERF model was peer reviewed prior to the 2002 PRA Revision 2 using Boiling Water Reactor (BWR) Owners Group (BWROG) process. The review team used the "BWROG PSA Peer Review Certification Implementation Guidelines," Revision 3, January 1997. Facts and Observation sheets documented the certification team's insights and potential level of significance. All of the 'A' priority PRA peer review comments have been addressed and incorporated into the GGNS PRA model as appropriate. All of the 'B' priority comments have been addressed except for one documentation item related to the internal flood modeling.Following the Integration and Quantification Task of the Rev. 2 and Rev. 3 model updates, an expert panel of GGNS personnel met to review model quantification results (top 100 cutsets). Various departments (Training, Operations, Engineering and Nuclear Safety) within the GGNS organization were invited to participate. Each of the top 100 cutsets was reviewed individually. In addition, cutsets from accident sequences representing approximately 99 percent of the total core damage frequency were also reviewed if there were no cutsets from these sequences in the top 100. The focus of the review was to identify poor assumptions, over-simplifications, incorrect credit for human actions, sequence timing errors, system modeling errors, and incorrect event probabilities. The reviews resulted in modifications to the model and to the credit given for human actions.As part of the EPU Level 2 PRA model development, an expert panel review of the preliminary cutsets was performed. The expert panel consisted of members of the Entergy PRA staff and the contractor staff who were developing the Level 2 portion of the PRA model. The purpose of this expert panel review was to provide an assessment of a preliminary Level 2 PRA model and its resulting cutsets. This feedback was then used to correct the model and ensure that the final model incorporated the lessons learned from the initial model development. E.1.5The MACCS2 Model-Level 3 AnalysisE.1.5.1Introduction SAMA evaluation relies on Level 3 PRA results to measure the effects of potential plant modifications. A Level 3 PRA model using version 1.13.1 of the MELCOR Accident RPV Rupturenot modelednot modeled0.4%0.3%Loss of Service Transformernot modelednot modeled6.5%7.1%1.Special initiators include loss of AC bus, DC bus, service water, closed cooling water, or instrument air.Contribution to CDF Changes in PRA Models (Continued)Contributing Initiator GroupR1R2R3R3 EPU Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-72Consequences Code System Version 2 (MACCS2) [

E.1-2] was created for GGNS. This model, which requires detailed site-specific meteorological, population, and economic data, estimates the consequences in terms of population dose and offsite economic cost. Risks in terms of population dose risk (PDR) and offsite economic cost risk (OECR) were also estimated in this analysis. Risk is defined as the product of consequence and frequency of an accidental release.This analysis considers a base case and two sensitivity cases to account for variations in data and assumptions for postulated internal events. The base case uses estimated time and speed for evacuation. Sensitivity case 1 is the base case with delayed evacuation. Sensitivity case 2 is the base case with lower evacuation speed.PDR was estimated by summing over all releases the product of population dose and frequency for each accidental release. Similarly, OECR was estimated by summing over all releases the product of offsite economic cost and frequency for each accidental release. Offsite economic cost includes costs that could be incurred during the emergency response phase and costs that could be incurred through long-term protective actions.E.1.5.2InputThe following sections describe the site-specific input parameters used to obtain the off-site dose and economic impacts for cost-benefit analyses.E.1.5.2.1Projected Total Population by Spatial ElementThe total population within a 50-mile radius of GGNS was estimated for the year 2044. Areal weighting was used to transfer the 2044 projected total population from source areas (county) to target areas (spatial elements) by converting county population to a density measure (e.g., number of people in county/acre) and multiplying this density by the area that county has in a spatial element. For spatial elements comprised of elements of more than one county, individual county densities were multiplied by areas of each county in a spatial element and summed. For counties with declining populations, the US Census 2000 values were used to provide a conservative estimate. Louisiana and Mississippi state tourism data was used to calculate a transient to permanent population ratio to increase each county's projected population to account for visitors. Total projected population of the 50-mile zone of analysis is 359,039, and the distribution of the 2044 total population is summarized in TableE.1-11

.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-73Table E.1-11Estimated Population Distribution within a 50-Mile RadiusWind Direction0 to 10miles11 to 20miles21 to 30miles31 to 40miles41 to 50milesTotalN25916597656652,205NNE275,44734,0954,35479744,720NE1221,9384,5385,2013,91315,712ENE2522393,4954,78470,71079,480 E4046561,5614,75016,60323,974ESE1,3201,0433546,93110,84920,497SE3,4361,3717393,40728,41837,371 SSE6021,1585843,4722,6688,484S1242,3534,8811,9491,38310,690SSW7361,4262,44529,7324,60638,945 SW 2503751,49314,6463,38720,151WSW 881,7402971,7812,1626,068W 1033163514,5053,0808,355 WNW 202,4092634,45111,26018,403NW 12136572,2396,3328,776NNW 39411,5677142,83015,208Totals 7,52420,79267,37993,681169,663359,039 Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-74E.1.5.2.2Land FractionThe National Hydrography Dataset for the watersheds within the 50-mile radius area was used to calculate the extent of land and surface water coverage. Calculated values ranged from 0.00 to 1.00. A value of 1.00 indicates the spatial element area is all land, with no significant surface water.E.1.5.2.3Watershed ClassWatershed Index is defined by MACCS2 as areas drained by rivers (Class 1) or large water bodies (Class 2). Class 2 is intended only for use with a very large lake, similar in size to Lake Michigan. For GGNS, a watershed index of 1 (drained by rivers) was used for all spatial elements.E.1.5.2.4Regional Economic DataRegion IndexEach spatial element was assigned to an economic region, defined in this report as a county. When a spatial element was comprised of more than one county, it was assigned to the county that had the most area in that spatial element. Two parishes in Louisiana (Caldwell and West Carroll) and seven counties in Mississippi (Amite, Madison, Rankin, Sharkey, Simpson, Wilkinson, and Yazoo) were not assigned due to their small representation in any one spatial element.Regional Economic DataEconomic data was obtained from the US Census of Agriculture (USDA 2007) for 2007, Department of Commerce and Department of Labor Statistics.VALWF- Value of Farm WealthMACCS2 requires an average value of farm wealth (dollars/hectare) for the 50-mile radius area around GGNS. The county-level farmland property value was used as a basis for deriving this value. VALWF is $4,787.34/hectare.VALWNF- Value of Non-Farm WealthMACCS2 also requires an average value of non-farm wealth. The county-level non-farm property value was used as a basis for deriving this value. VALWNF is $97,224.14/person.Other economic parameters and their values are shown below. The values were calculated using average U.S. Consumer Prices Indices. A proportional factor of 1.9 was developed using the December 1987 CPI (113.6) and the December 2010 CPI (218.056). This CPI factor was applied to the previously recommended values of the following parameters to represent current values.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-75E.1.5.2.5Agriculture DataThe source of regional crop information is the 2007 United States Census of Agriculture. The crops listed for each county within the 50-mile area were summed and mapped into the seven MACCS2 crop categories.E.1.5.2.6Meteorological DataThe MACCS2 model requires meteorological data for wind speed, wind direction, atmospheric stability, accumulated precipitation, and atmospheric mixing heights. The required data was obtained from the GGNS meteorological monitoring system and regional National Weather Service stations.Site-Specific DataMeteorological data collected at the site from calendar years 2005 through 2009 were compiled for the MACCS2 input file. Missing data for parameters of interest were estimated using data substitution methods. These methods include substitution of missing data with valid data from the previous hour and substitution of valid data collected from other elevations on the meteorological tower. The 2009 data resulted in the highest release quantities and was therefore used to perform the base case analysis and sensitivity cases.Regional Mixing Height DataMixing height is defined as the height of the atmosphere above ground level within which a released contaminant will become mixed (from turbulence) within approximately one hour. GGNS mixing height data were estimated using the ground level and upper-air data from the National Weather Service.VariableDescriptionValueCHEVACST001Daily cost for a person who has been evacuated ($/person-day)51.3CHPOPCST001Population relocation cost ($/person)9500CHRELCST001Daily cost for a person who is relocated ($/person-day)51.3 CHCDFRM001Cost of farm decontamination for the various levels of decontamination ($/hectare)1068.752375CHCDNFRM001Cost of non-farm decontamination for the various levels of decontamination ($/person)570015200CHDLBCST001Average cost of decontamination labor ($/person-year)66500 DPRATEProperty depreciation rate (per year)0.2DSRATEInvestment rate of return (per year)0.12 Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-76E.1.5.2.7Emergency Response AssumptionsA detailed analysis of evacuation scenarios in the 10-mile emergency planning zone (EPZ) were addressed in the GGNS evacuation travel time estimate study for both the Mississippi side (Claiborne and Warren counties) and the Louisiana side (Tensas Parish) of the Mississippi River

[E.1-3]. These studies, conducted from August through December 2006, provide an analysis of the range and variation of public reaction to the evacuation notification process. This is the most recent report available and is still valid because the population in the two counties and a parish with land in the 10-mile EPZ has been in decline since the studies were conducted.Evacuation Delay TimeThe estimates for the general public were based on the following evacuation components: notification, preparation to depart, and actual evacuation. The evacuation study concluded that 100 percent of the general public would be prepared to begin an evacuation within 195 minutes from activation of the evacuation notification process. This includes 50 minutes for notification and 145 minutes for the population to get ready to leave, for a total delay time of 195 minutes.Evacuation SpeedThe evacuation travel time studies concluded that in the worst case the general public within the 10-mile EPZ could be evacuated in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 10 minutes (250 minutes) from issuance of an order to evacuate for 100 percent of the population. Total evacuation time includes the delay time discussed above. Since 195 minutes of this is the delay time, the worst case transit time is 55 minutes. The longest travel times were required for evacuation scenarios occurring on mid-week days in adverse weather (rain).Evacuation travel speed is calculated by dividing the distance traveled by the time required to evacuate 100 percent of the total population. Since the maximum travel distance out of the EPZ is 10 miles, the general public transit speed is 10 mi / 55 min = 10.9 mph (4.87m/s).E.1.5.2.8Core InventoryThe GGNS core inventory is shown in TableE.1-12. These values are based on ORIGEN 2.1 evaluations supporting the EPU to 115% (4408 MWt) of the original licensed thermal power.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-77Table E.1-12Estimated GGNS Core Inventory (Becquerels)(1)NuclideInventoryNuclideInventoryCo-584.22E+16Te-131m6.59E+17 Co-607.29E+16Te-1326.40E+18Kr-855.81E+16I-1314.51E+18Kr-85m1.19E+18I-1326.51E+18Kr-872.28E+18I-1339.18E+18Kr-883.20E+18I-1341.01E+19Rb-861.12E+16I-1358.58E+18Sr-894.33E+18Xe-1338.81E+18Sr-904.63E+17Xe-1353.12E+18Sr-915.40E+18Cs-1341.04E+18 Sr-925.85E+18Cs-1363.35E+17Y-904.92E+17Cs-1376.18E+17Y-915.59E+18Ba-1398.18E+18 Y-925.88E+18Ba-1407.92E+18Y-936.77E+18La-1408.40E+18Zr-957.99E+18La-1417.47E+18 Zr-978.25E+18La-1427.22E+18Nb-958.03E+18Ce-1417.51E+18Mo-998.55E+18Ce-1436.96E+18Tc-99m7.44E+18Ce-1446.14E+18Ru-1037.14E+18Pr-1436.73E+18Ru-1055.03E+18Nd-1473.00E+18 Ru-1062.76E+18Np-2399.32E+19Rh-1054.74E+18Pu-2381.89E+16Sb-1275.00E+17Pu-2391.91E+15Sb-1291.48E+18Pu-2402.58E+15 Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-78E.1.5.2.9Source TermsEleven release categories, corresponding to internal event sequences, were part of the MACCS2 input. Section E.1.2.2.6 provides details of the source terms for postulated internal events. A linear release rate was assumed between the time the release started and the time the release ended.E.1.5.3RESULTS Risk estimates for one base case and two sensitivity cases were analyzed with MACCS2. Sensitivity Case 1 assumes an evacuation time delay that is increased from 3.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> (base) to 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Sensitivity Case 2 assumes a lower average evacuation speed; the speed was reduced from 4.87 m/s (base) to 2.435 m/s. TableE.1-13 shows estimated base case mean risk values for each release mode. The estimated mean values of PDR and offsite OECR for GGNS are 0.486 person-rem/yr and

$1,244/yr, respectively.Te-1274.96E+17Pu-2418.44E+17Te-127m6.70E+16Am-2419.44E+14Te-1291.45E+18Cm-2422.50E+17Te-129m2.16E+17Cm-2441.58E+161.From GGNS specific data for a power level of 4408 MWth [

E.1-2].Table E.1-12 (Continued)Estimated GGNS Core Inventory (Becquerels)(1)NuclideInventoryNuclideInventory Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-79Results of sensitivity analyses indicate that a delayed evacuation or a lower evacuation speed would not have any significant effects on the offsite consequences or risks determined in this study. TableE.1-14 summarizes offsite consequences in terms of population dose (person-sv) and offsite economic cost ($) for the base case and the sensitivity cases. Comparison of the consequences indicates a deviation of less than 1% between the base case and the sensitivity case results.Table E.1-13Base Case Mean PDR and OECR Values for Postulated Internal EventsCharacteristics of Release Mode (1)Population Dose Offsite Economic CostPopulation Dose Risk (PDR) Offsite Economic Cost Risk (OECR)IDFrequency (per year)(person-sv)(1)person-rem($)(person-rem/yr)(2)$/yrH/L8.73E-084.12E+034.12E+051.04E+093.60E-029.08E+01H/E1.05E-072.29E+032.29E+053.13E+082.41E-023.29E+01H/I1.23E-085.10E+035.10E+051.37E+096.26E-031.68E+01M/E3.49E-074.66E+034.66E+051.31E+091.63E-014.57E+02M/I1.73E-076.70E+036.70E+051.81E+091.16E-013.14E+02M/L2.71E-073.86E+033.86E+051.04E+091.05E-012.82E+02L/E4.04E-099.92E+029.92E+047.32E+074.00E-042.95E-01L/I3.34E-083.26E+033.26E+057.48E+081.09E-022.50E+01L/L1.32E-071.75E+031.75E+051.66E+082.30E-022.18E+01LL/E2.00E-093.62E+003.62E+024.63E+057.24E-079.26E-04LL/I2.11E-091.80E+001.80E+024.59E+053.80E-079.68E-04LL/L6.83E-092.90E+032.90E+054.81E+081.98E-033.28E+00Totals 4.86E-011.24E+031.Conversion Factor: 1 sv = 100 rem.2.Value is the product of the release mode frequency and the population dose.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-80Table E.1-14Summary of Offsite Consequence Results for Sensitivity ResultsPopulation Dose (person-sv)(1)Offsite Economic Cost ($)Release Mode BaseLonger Time for Evacuation Slower Speed of Evacuation BaseLonger Time for Evacuation Slower Speed of EvacuationH/L4.12E+034.12E+034.12E+031.04E+091.04E+091.04E+09H/E2.29E+032.30E+032.29E+033.13E+083.13E+083.13E+08H/I5.10E+035.10E+035.10E+031.37E+091.37E+091.37E+09M/E4.66E+034.66E+034.66E+031.31E+091.31E+091.31E+09M/I6.70E+036.70E+036.70E+031.81E+091.81E+091.81E+09M/L3.86E+033.86E+033.86E+031.04E+091.04E+091.04E+09L/E9.92E+029.93E+029.96E+027.32E+077.32E+077.32E+07 L/I3.26E+033.27E+033.26E+037.48E+087.48E+087.48E+08L/L1.75E+031.75E+031.75E+031.66E+081.66E+081.66E+08LL/E3.62E+003.65E+003.63E+004.63E+054.63E+054.63E+05LL/I1.80E+001.83E+001.80E+004.59E+054.59E+054.59E+05LL/L2.90E+032.90E+032.90E+034.81E+084.81E+084.81E+08Total4.86E-014.86E-014.86E-011.24E+031.24E+031.24E+03person-rem/

yrperson-rem/

yrperson-rem/

yr$/yr$/yr$/yr1.Conversion Factor: 1 sv = 100 rem.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-81E.1.6ReferencesE.1-1NEI 05-01, Severe Accident Mitigation Alternat ives (SAMA) Analysis Guidance Document, November 2005, Revision A.E.1-2CALC-OC-N1000-10002, "GGNS Level 3 Probabilistic Safety Analysis (PSA) Model" Rev. 0.E.1-3Grand Gulf Nuclear Station Development of Evacuation Time Estimates, KLD Associates, Inc., September 2007.E.1-4PRA-GG-01-003, Grand Gulf Power Station Detailed Level 2 Analysis, Revision 0, August 2010.E.1-5GGNS PRA-GG-01-001, "GGNS Level-1 Model Revision 3 PSA Summary Report," Rev. 2.E.1-6Intentionally Left Blank E.1-7Kaiser, "The Implications of Reduced Source Terms for Ex-Plant Consequence Modeling," Executive Conference on the Ramifications of the Source Term (Charleston, SC), March 12, 1985.E.1-8USNRC, NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, 1991.E.1-9GGNS94-0053 IPEEE, "Internal Plant Examination of External Events Seismic Margins," Revision 0.E.1-10USNRC, NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, June 1991.E.1-11EPRI NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," August 1991.E.1-12EPRI Fire PRA Implementation Guide, prepared by Science Applications International Corporation for Electric Power Research Institute, January 1994.E.1-13 Cottle, W. T. to USNRC, "GGNS Respon se to Generic Letter 88-20, 'Individual Plant Examination for Severe Accidents Vulnerabilities,'" Correspondence No. GNRO-92/00157, letter dated December 23, 1992.E.1-14GGNS-97-0014, "GGNS PRA Update Summary and Results Report," July 30, 1997.E.1-15GGNS95-00041, "Internal Plant Examination of External Events Fire," October 1996.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.1-82E.1-16USNRC to R. Hutchinson (GGNS), "Generic Letter 88-20, Individual Plant Examination (IPE) - Internal Events - Grand Gulf Nuclear Station (TAC M74415)," Correspondence No. GNRI-96/00067, letter dated March 7, 1996.E.1-17GGNS Calculation No. XC-N1111-01007, "GGNS Level 1 PSA," Revision 2, October 17, 2002.E.1-18GGNS Calculation No. PRA-GG-09-001, "Identification of Risk Implications due to Extended Power Uprate at Grand Gulf," May 2010.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal StageATTACHMENT E.2 EVALUATION OF GGNS SAMA CANDIDATES Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-1E.2 EVALUATION OF GGNS SAMA CANDIDATESThis section describes the generation of the initial list of potential SAMA candidates, screening methods, and the analysis of the remaining SAMA candidates.E.2.1SAMA List CompilationCandidate SAMAs are defined as potential enhancements to the plant design, operating procedures, inspection programs, or maintenance programs that have the potential to reduce the severe accident risk of GGNS. These SAMAs can be characterized as either hardware (i.e.,

physical modification of plant structure, systems, and components) or non-hardware enhancements (i.e., operation, maintenance programs, and procedure changes), or a combination of the two. The candidate SAMAs considered for GGNS encompass both hardware and non-hardware enhancements.A list of SAMA candidates was developed by reviewing industry documents and considering other plant-specific enhancements not identified in published industry documents. Since GGNS is a BWR, considerable attention was paid to the SAMA candidates from SAMA analyses for other BWR plants. Industry documents reviewed include the following. *NEI 05-01, Severe Accident Mitigation Alternatives Analysis [

E.2-1] *James A. FitzPatrick Nuclear Power Plant SAMA Analysis [

E.2-2]*Vermont Yankee Nuclear Power Station SAMA Analysis [E.2-3]*Pilgrim Nuclear Power Station SAMA Analysis [

E.2-4]*Oyster Creek Nuclear Generating Station SAMA Analysis [

E.2-5]*Monticello Nuclear Generating Plant SAMA Analysis [

E.2-6]*Brunswick Steam Electric Plant, Units 1 and 2 SAMA Analysis [

E.2-7]*NUREG-1742, Perspectives Gained from the Indivi dual Plant Examinat ion of External Events (IPEEE) Program

[E.2-8]*Duane Arnold Energy Center [E.2-11]*Susquehanna Steam Electric Station, Units 1 and 2 [

E.2-10]*Cooper Nuclear Station, Unit 1 [E.2-9]In addition to SAMA candidates from review of industry documents, additional SAMA candidates were obtained from plant-specific sources, such as the GGNS IPE [

E.2-18] and the GGNS IPEEE [E.2-13 , E.2-14 , E.2-15 , E.2-16 , E.2-17]. In the IPE and IPEEE several enhancements related to severe accident insights were recommended and implemented. These enhancements are included in the comprehensive list of Phase I SAMA candidates as 226 through 245 (see TableE.2-1). The current GGNS PSA levels1 and 2 models were also used to identify plant-specific modifications for inclusion in the comprehensive list of SAMA candidates. The risk significant events from the current PSA model were reviewed for sim ilar failure modes and effects that could be addressed through a potential enhancement to the plant. The correlation between SAMAs and the risk significant terms are listed in Tables E.1-2 and E.1-4.The comprehensive list of 249 candidate SAMAs considered for implementation at GGNS is provided in onsite documentation [

E.2-21].

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-2E.2.2Qualitative Screening of SAMA Candidates (Phase I)The purpose of the preliminary SAMA screening was to eliminate from further consideration enhancements that were not viable for implementation at GGNS. Potential SAMA candidates were screened out if they modified features not applicable to GGNS, if they had already been implemented at GGNS, or if they were similar in nature and could be combined with another SAMA candidate to develop a more comprehensive or plant-specific SAMA candidate. During this process, 60 of the Phase I SAMA candidates were screened out because they were not applicable to GGNS, 28 of the Phase I SAMA candidates were screened out because they were similar in nature and could be combined with another SAMA candidate, and 98 of the Phase I SAMA candidates were screened out because they had already been implemented at GGNS, leaving 63 SAMA candidates for further analysis. The final screening process involved identifying and eliminating those items whose implementation cost would exceed their benefit as described below. TableE.2-2 provides a description of each of the 63 Phase II SAMA candidates.E.2.3Final Screening and Cost Benefit Evaluation of SAMA Candidates (Phase II)A cost/benefit analysis was performed on each of the remaining SAMA candidates. If the implementation cost of a SAMA candidate was determined to be greater than the potential benefit (i.e., there was a negative net value) the SAMA candidate was considered not to be cost beneficial and was not retained as a potential enhancement.The expected cost of implementation of each SAMA was established from existing estimates of similar modifications. Most of the cost estimates were developed from similar modifications considered in previously performed SAMAs. In particular, these cost-estimates were derived from the following sources.*Pilgrim Nuclear Power Station [

E.2-4]*Hope Creek [

E.2-12]*Columbia Generating Station [

E.2-19]*Cooper Nuclear Station [E.2-9]*Duane Arnold Energy Center [E.2-11]The benefit of implementing a SAMA candidate was estimated in terms of averted consequences by altering the base case PSA model to reflect the maximum benefit of the improvement and re-quantifying the PDS frequency with a truncation of 1E-12. The benefit was estimated by calculating the arithmetic difference between the total estimated costs associated with the four impact areas for the baseline plant design and the total estimated impact area costs for the enhanced plant design (following implementation of the SAMA candidate).Values for avoided public and occupational health risk were converted to a monetary equivalent (dollars) via application of the Regulatory Analysis Technical Evaluation Handbook

[E.2-20] conversion factor of $2,000 per person-rem and discounted to present value. Values for avoided off-site economic costs were also discounted to present value.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-3As this analysis focuses on establishing the economic viability of potential plant enhancement when compared to attainable benefit, detailed cost estimates often were not required to make informed decisions regarding the economic viability of a particular modification. The implementation costs for several of the SAMA candidates were clearly in excess of the attainable benefit estimated from a particular analysis case. Nonetheless, the cost of each SAMA candidate was conceptually estimated to the point where conclusions regarding the economic viability of the proposed modification could be adequately gauged.Based on a review of previous SAMA evaluations and an evaluation of expected implementation costs at GGNS, the following estimated cost ranges for each type of proposed SAMA implementation were used.Detailed cost estimates were based on the engineering judgment of project engineers experienced in performing design changes at the facility. The detailed cost estimates considered engineering, labor, materials, and support fu nctions such as planning, scheduling, health physics, quality assurance, security, safety, and firewatch. The estimates included a 20% contingency on the design and installation costs but did not account for inflation, replacement power during extended outages necessary for SAMA implementation, or increased maintenance or operation costs following SAMA implementation. The cost benefit comparison and disposition of each of the 63 Phase II SAMA candidates is presented in TableE.2-2

.Bounding evaluations (or analysis cases) were performed to address s pecific SAMA candidates or groups of similar SAMA candidates. These analysis cases overestimated the benefit and thus were conservative calculations. For example, one SAMA candidate suggested installing digital large break LOCA protection; the bounding calculation estimated the benefit of this improvement by total elimination of risk due to large break LOCA (see analysis of Phase II SAMA 56 in Table E.2-2). This calculation obviously overestimated the benefit, but if the inflated benefit indicated that the SAMA candidate was not cost beneficial, then the purpose of the analysis was satisfied.A description of the analysis cases used in the evaluation follows.

Case 1:DC PowerThis analysis case was used to evaluate the change in plant risk from provide additional DC battery capacity. A bounding analysis was performed by eliminating station blackout cutsets from Type of Change Estimated Cost RangeProcedural only$25K-$50KProcedural change with engineering or training required$50K-$200KProcedural change with engineering and testing/training required$200K-$300KHardware modification$100K to > $1000K Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-4the PSA model [basic events ZSBO and ZT1B were set to zero], which resulted in an internal and external benefit (with uncertainty) of approximately $346,968. This analysis case was used to model the benefit of Phase II SAMAs 1, 2, 11, 12, and 15.Case 2:Improve Charger ReliabilityThis SAMA analysis case was used to evaluate the change in plant risk from improving the diversity of the DC battery charging capability by adding an additional battery charger or providing a means to lower battery charger failure. A bounding analysis was performed by setting the failure of chargers contribution to zero in the leve l 1 PSA model. The following basic events were removed from the model:This resulted in an internal and external benefit (with uncertainty) of approximately $40,793. This analysis case was used to model the benefit of Phase II SAMAs 3 and 13.Case 3:Add DC System Cross-TiesThis analysis case was used to evaluate the change in plant risk from providing DC bus cross-ties. A bounding analysis was perf ormed by eliminating failure of DC power gates in the PSA model (with the following gates removed from the model: 11DA-001, 11DA-001-SBO, 11DA-001T, 11DA-001X, 11DA-001Y, 11DA-001Z, 11DB-001, 11DB-001-SBO, 11DB-001T, 11DB-001X, and 11DB-001Z), which resulted in an internal and external benefit (with uncertainty) of approximately

$219,169. This analysis case was used to model the benefit of Phase II SAMA 4.11DA-007-D 11DA-008-D 11DB-007-E 11DB-008-E11DC-007-F 11DC-008-F 11DD-007-X 11DD-008-X11DE-007-X 11DE-008-X L21-CO-CB11A02-D L21-CO-CB11A03-D L21-CO-CB11B02-E L21-CO-CB11B03-E L21-CO-CB11D02-X L21-CO-CB11D03-XL21-CO-CB11E02-X L21-CO-CB11E03-X L51-LP-BC-1A4-D L51-LP-BC-1A5-DL51-LP-BC-1B4-E L51-LP-BC-1B5-E L51-LP-BC-1D4-X L51-LP-BC-1D5-X L51-LP-BC-1E4-X L51-LP-BC-1E5-X L51-MA-BC-1A4-D L51-MA-BC-1A5-DL51-MA-BC-1B4-E L51-MA-BC-1B5-E L51-MA-BC-1D4-X L51-MA-BC-1D5-XL51-MA-BC-1E4-X L51-MA-BC-1E5-X P81-CO-CB11C02-F P81-CO-CB11C03-F P81-CO-CB70104-F P81-FO-HE1C5-F P81-LP-BC-1C4-F P81-LP-BC-1C5-FP81-MA-BC-1C4-F P81-MA-BC-1C5-F R20-CF-CB-BKR R20-CO-CB15102-XR20-CO-CB15202-X R20-CO-CB15306-D R20-CO-CB15602-D R20-CO-CB16102-XR20-CO-CB16202-X R20-CO-CB16306-E R20-CO-CB16602-E R20-CO-CB31116-F Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-5Case 4:Increase Availability of On-Site AC PowerThis analysis case was used to evaluate the change in plant risk from improving the backup sources for the Vital AC buses 15AA, 16AB, and 17AC. A bounding analysis was performed by eliminating failure of DG11, DG12, and DG13 to their AC buses (15AA, 16AB, and 17AC, respectively) in the Level 1 model (with the following gates set to zero: DG11-001L, DG11-001T, DG11-001X, DG11-001X-HPCS, DG11-001X-ONSP, DG11-001XP, DG11-001XZ, DG12-001L, DG12-001T, DG12-001X, DG12-001XP, DG12-001XZ, DG13-001N, DG13-001X, DG11-06, DG12-06, SBO1-DG13-001X, and SBO2-DG13-001X), which resulted in an internal and external benefit (with uncertainty) of approximately $448,189. This analysis case was used to model the benefit of Phase II SAMAs 5 and 8.Case 5:Improve AC PowerThis analysis case was used to evaluate the change in plant risk from improving the 4.16-kV bus cross-tie ability. A bounding analysis was performed by eliminating the loss of the 4.16-kV buses in the PSA model [with the following gates removed from the model: 15AA-001, 15AA-001D, 15AA-001-HPCS, 15AA-001L, 15AA-001P, 15AA-001T, 15AA-001U, 15AA-001Z, 16AB-001, 16AB-001D, 16AB-001-HPCS, 16AB-001L, 16AB-001ONSP, 16AB-001P, 16AB-001T, 16AB-001U, 16AB-001Z, 17AC-001, 17AC-001-DGX, and 17AC-001N], which resulted in an internal and external benefit (with uncertainty) of approximately $532,571. This analysis case was used to model the benefit of Phase II SAMAs 6 and 17.Case 6:Reduce Loss of Off-Site Power During Severe WeatherThis SAMA analysis evaluated the change in plant risk from installing an additional buried off-site power source. A bounding analysis was performed by removing LOSP due to severe weather from the LOSP initiating event frequencies [%T1 and %T1P were multiplied by 19/24]. This resulted in an internal and external benefit (with uncertainty) of approximately $78,261. This analysis case was used to model the benefit of Phase II SAMA 7. Case 7:Provide Backup Emergency Diesel Generator (EDG) CoolingThis analysis case was used to evaluate the change in plant risk from increasing EDG reliability by adding a backup source of diesel cooling. A bounding analysis was performed by eliminating failure of SW cooling to the EDGs [the following gates were eliminated: DGA-001L, DGA-001T, DGA-001X, DGA-001X-HPCS, DGA-001X-ONSP, DGA-001XP, DGA-001XZ, DGB-001L, DGB-001T, DGB-001X, DGB-001XP, DGB-001XZ, DGC-001N, and DGC-001X], which resulted in an internal and external benefit (with uncertainty) of approximately $49,545. This analysis case was used to model the benefit of Phase II SAMAs 9 and 10.Case 8:Increase EDG ReliabilityThis analysis case was used to evaluate the change in plant risk from providing a portable EDG fuel oil transfer pump. A bounding analysis was performed by eliminating failure of EDGs to run in the PSA model [the following basic events ere set to zero: P75-FR-DG-DG11-A, P75-FR-DG-Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-6DG12-B, P75-CF-3DGR-Z, and P75-CF-DGR-Z], which resulted in an internal and external benefit (with uncertainty) of approximately $91,044. This analysis case was used to model the benefit of Phase II SAMA 14.Case 9:Improve DG reliabilityThis analysis case was used to evaluate the change in plant risk from providing a diverse swing diesel generator air start compressor. A bounding analysis was performed by eliminating the common cause failure (CCF) contribution of failure to start EDGs in the PSA model [the following CCF events were set to zero: P75-CF-3DGS-Z and P75-CF-DGS-Z], which resulted in an internal and external benefit (with uncertainty) of approximately $6,542. This analysis case was used to model the benefit of Phase II SAMA 16.Case 10:Reduce Plant-Centered Loss of Off-Site PowerThis analysis case was used to evaluate the change in plant risk from protecting transformers from failure. A bounding analysis was performed by removing the initiating contribution of plant and switchyard centered events in the PSA mo del. The LOSP notebook does not discriminate transformer failures between switchyard-centered or plant-centered so all plant-centered and switchyard-centered LOSP events were removed from the LOSP frequency [%T1 and %T1P were multiplied by 9/24], which resulted in an internal and external benefit (with uncertainty) of approximately $229,668. This analysis case was used to model the benefit of Phase II SAMA 18.Case 11:Redundant Power to Torus Hard Pipe Vent (THPV) ValvesThis analysis case was used to evaluate the change in plant risk from providing redundant power to the direct torus vent valves. A bounding analysis was performed by eliminating failure of power to containment vents in the PSA model, which resulted in an internal and external benefit (with uncertainty) of approximately $32,297. This analysis case was used to model the benefit of Phase II SAMA 19.Specifically, the following gates were set to zero or removed:*15P21-001PROB 0*16P41-001PROB 0

  • 1DA1-001 deleted from M41-002, M41-002X, and VC-L2-AC-POWER*1DB1-001 deleted from M41-002, M41-002X, and VC-L2-AC-POWER Case 12:High Pressure Injection SystemThis analysis case evaluated the change in plant risk from plant modifications that would increase the availability of high pressure core spray (installing a high pressure injection system independent of AC power or a passive high pressure core injection system). A bounding analysis was performed by eliminating failure of HPCS in the PSA model [gates U1, U1-RX, and U1-SI were removed from the model], which resulted in an internal and external benefit (with Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-7uncertainty) of approximately $1,784,736. This analysis case was used to model the benefit of Phase II SAMAs 20 and 61.Case 13:Extend RCIC Operation This analysis case was used to evaluate the change in plant risk from raising the RCIC back pressure trip setpoint. A bounding analysis was performed by eliminating failure of trip due to

pressure in the PSA model [gate E51-400 was set to zero], which resulted in an internal and external benefit (with uncertainty) of approximately $30,093. This analysis case was used to model the benefit of Phase II SAMA 21.Case 14:Improve ADS SystemThis analysis case was used to evaluate the change in plant risk from modifying the automatic depressurization system (ADS) components to improve reliability by adding larger accumulators. A bounding analysis was performed by eliminating failure of ADS valves in the PSA model [gates B21-001B1 and B21-003 were set to zero], which resulted in an internal and external benefit (with uncertainty) of approximately $897,317. This analysis case was used to model the benefit of Phase II SAMA 22.Case 15:Improve ADS SignalsThis analysis case was used to evaluate the change in plant risk from adding signals to open safety relief valves automatically in an MSIV closure transient. A bounding analysis was performed by eliminating failure of the SRV to open in the PSA model [the following gates were set to zero: OP-DEPRESS-OP1, B21-001B1, B21-001A, B21-006 and ba sic event B21-CF-SF-K], which resulted in an internal and external benefit (with uncertainty) of approximately $388,150. This analysis case was used to model the benefit of Phase II SAMA 23.Case 16:Low Pressure Injection SystemThis analysis case was used to evaluate the change in plant risk from adding a diverse low pressure injection system. A bounding analysis was performed by eliminating failure of LPCI and low pressure core spray (LPCS) in the PSA model [the following gates were set to zero: V2, V2-RX, V2-SI, V3, V3-RX, V3-SI, and V3-SBO], which resulted in an internal and external benefit (with uncertainty) of approximately $689,896. This analysis case was used to model the benefit of Phase II SAMA 24.Case 17:Emergency Core Cooling System (ECCS) Low Pressure InterlockThis analysis case was used to evaluate the change in plant risk from installing a bypass switch to allow operators to bypass the low reactor pressure interlock circuitry that inhibits opening the LPCI or core spray injection valves following sensor or logic failures that prevent all low pressure injection valves from opening. A bounding analysis was performed by eliminating ECCS permissives and interlock failure in the PSA model [the following gates were set to zero: E12-110, E12-190, B21-012A, B21-013A, B21-026A, and B21-027A], which resulted in an internal Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-8and external benefit (with uncertainty) of approximately $30,093. This analysis case was used to model the benefit of Phase II SAMA 25.Case 18:RHR Heat ExchangersThis analysis case was used to evaluate the change in plant risk from implementing modifications to allow manual alignment of the fire water system to RHR heat exchangers. A bounding analysis was performed by eliminating failure of SSW to provide cooling to the RHR heat exchangers [the following gates were removed from the model: P41-RHRHXA-SBO, P41-RHRHXB-SBO, P41-RHRHXA and P41-RHRHXB], which resulted in an internal and external benefit (with uncertainty) of approximately $615,669. This analysis case was used to model the benefit of Phase II SAMA 26.Case 19:Emergency Service Water System ReliabilityThis analysis case was used to evaluate the change in plant risk from installing an additional service water pump. A bounding analysis was performed by eliminating failure of service water pumps in the PSA model [the following basic events were set to zero: P41-CF-MCP001R-R, P41-CF-MCP001S-R, P41-CF-MVDISNA-R, P41-CF-MVDISNB-R, P41-CF-MVDISNC-R, P41-CF-MVF001AB, P41-CF-MV-F001AB, P41-CF-MVF005AB, and P41-CF-ST-SUCT-R], which resulted in an internal and external benefit (with uncertainty) of approximately $113,708. This analysis case was used to model the benefit of Phase II SAMA 27.Case 20:Main Feedwater System ReliabilityThis analysis case was used to evaluate the change in plant risk from installing a motor-driven feedwater pump. A bounding analysis was performed by setting failure to inject from feedwater to zero in the PSA model [gate N21-002 was set to zero], which resulted in an internal and external benefit (with uncertainty) of approximately $486,149. This analysis case was used to model the benefit of Phase II SAMA 28.Case 21:Increase Availability of Room CoolingThis analysis case was used to evaluate the change in plant risk from providing a redundant HVAC train to rooms dependent on room cooling. A bounding analysis was performed by eliminating failure of room cooling to the safeguard switchgear battery rooms, standby service water pump rooms, LPCS pump rooms, and HPCS pump rooms in the PSA model [the following gates were set to zero: T51-060, Z77-300, T51-080, HVC-1000XP, HVC-1000XZ, HVC-1000-HPCS, HVC-1000X-HPCS, HVC-1000X-ONSP, HVC-1000X-SBO, and HVC-2000X], which resulted in an internal and external benefit (with uncertainty) of approximately $526,200. This analysis case was used to model the benefit of Phase II SAMA 29.Case 22:Increase Availability of the DG System through HVAC ImprovementsThis analysis case was used to evaluate the change in plant risk from enhancing diesel generator room cooling. A bounding analysis was performed by eliminating failure of cooling of Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-9three diesel generator rooms in the PSA model [gates HVC-001X, HVC-010X, and HVC-020X were set to zero]. This resulted in an internal and external benefit (with uncertainty) of approximately $227,963. This analysis case was used to model the benefit of Phase II SAMAs 30, 32, and 33.Case 23:Increase Reliability of HPCI and RCIC Room CoolingThis analysis case was used to evaluate the change in plant risk from creating the ability to switch HPCI and RCIC room fan power supply to DC in an SBO event. Since RCIC pump continued operation is not dependent on room cooling, a bounding analysis was performed by eliminating failure of power to the HPCS pump room cooler in the PSA model [gate 17B01-001 was removed from gate T51-080], which resulted in an internal and external benefit (with uncertainty) of approximately $30,093. This analysis case was used to model the benefit of Phase II SAMA 31.Case 24:Increase Reliability of Instrument AirThis analysis case was used to evaluate the change in plant risk from improving the reliability of the instrument air system. A bounding analysis was performed by eliminating failure of the instrument air system in the level 1 PSA model [the following gates were se t to zero: P53-001, P53-001AX, P53-001X, P53-101, P53-001A, P53-101X, P53-102, P53-102X, and initiator %TIA], which resulted in an internal and external benefit (with uncertainty) of approximately $413,527. This analysis case was used to model the benefit of Phase II SAMAs 34 and 35.Case 25:Backup Nitrogen to SRVThis analysis case was used to evaluate the change in plant risk from installing permanent nitrogen bottles as backup gas supply. A bounding analysis was performed by eliminating operator failure to install bottles in the PSA model [basic event B21-FO-HEBOTTLES was set to zero], which resulted in an internal and external benefit (with uncertainty) of approximately $121,841. This analysis case was used to model the benefit of Phase II SAMA 36.Case 26:Improve Availability of SRVs and MSIVsThis analysis case was used to evaluate the change in plant risk from improving SRV and MSIV pneumatic components. A bounding analysis was performed by eliminating failure of non-ADS SRVs in the PSA model [gate B21-004 and basic events B21-FO-HEDEP2-I and B21-CF-SF-K were set to zero], which resulted in an internal and external benefit (with uncertainty) of approximately $901,893. This analysis case was used to model the benefit of Phase II SAMA 37.Case 27:Improve Suppression Pool CoolingThis analysis case was used to evaluate the change in plant risk from installing an independent method of suppression pool cooling. This would allow the suppression pool to be an alternate cooling source for the RHR heat exchanger. A bounding analysis was performed by eliminating Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-10the failure of flow to the RHR heat exchangers in the PSA model [gates P41-RHRHXA, P41-RHRHXB, P41-RHRHXA-SBO, and P41-RHRHXB-SBO we re removed from the model], which resulted in an internal and external benefit (with uncertainty) of approximately $615,669. This analysis case was used to model the benefit of Phase II SAMA 38.Case 28:Increase Availability of Containment Heat RemovalThis analysis case was used to evaluate the change in plant risk from increasing the availability of containment heat removal. A bounding analysis was performed by eliminating failure of cooled flow through the injection line in the PSA model [gates E12-686, E12-686X, E12-686Y, E12-686Y-SBO, E12-686-SBO, E12-686X-SBO, E12-665, E12-665-SBO, E12-620, E12-620X, E12-620Y, E12-620-SBO, E12-620X-SBO, E12-620Y-SBO, E12-604, and E12-604-SBO were set to zero], which resulted in an internal and external benefit (with uncertainty) of approximately $865,312. This is similar to analysis case 29; however, the containment spray injection valves are not set to zero. This analysis case was used to model the benefit of Phase II SAMAs 39 and 41.Case 29. Decay Heat Removal Capability-Drywell SprayThis analysis case was used to evaluate the change in plant risk from improving drywell spray capability by installing a passive drywell spray system. Enhancements of decay heat removal capability decrease the probability of loss of containment heat removal. A bounding analysis was performed by setting the events for loss of RHR spray to zero in the PSA model [the following gates were set to zero: W3, W3X, #W3X, W3-SBO, W3X-SBO, W3Y, and W3Y-SBO], which resulted in an internal and external benefit (with uncertainty) of approximately $865,649. This analysis case was used to model the benefit of Phase II SAMA 40.Case 30:Increase Availability of the CSTThis analysis case was used to evaluate the change in plant risk from providing a means of replenishing CST water from the firewater, demineralized water, or service water system. A bounding analysis was performed by eliminating the CDF contribution from HPCS and RCIC suction [gates P11-F021 and E22-041 were set to zero], which resulted in an internal and external benefit (with uncertainty) of approximately $323,696. This analysis case was used to model the benefit of Phase II SAMA 42.Case 31:Filtered Vent to Increase Heat Removal Capacity for Non-ATWS Events This analysis case was used to evaluate the change in plant risk from installing a filtered containment vent. A bounding analysis was performed by reducing the baseline accident progression source terms by a factor of 2 (excluding noble gases) to reflect the additional filtered capability. Reducing the releases from the vent path resulted in an internal and external benefit (with uncertainty) of approximately $242,759. This analysis case was used to model the benefit of Phase II SAMA 43.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal StageE.2-11Case 32:Reduce Hydrogen IgnitionThis SAMA analysis case was used to evaluate the change in plant risk from installing a passive hydrogen control system or from providing post-accident containment inerting capability. A bounding analysis was performed by eliminating failure of hydrogen igniters in the PSA model [gate E61-001 was set to zero], which resulted in an internal and external benefit (with uncertainty) of approximately $427,365. This analysis case was used to model the benefit of Phase II SAMAs 44 and 45. Case 33: Controlled Containment VentingThis analysis case was used to evaluate the change in plant risk from enabling manual operation of all containment vent valves via local controls or from providing passive overpressure relief. A bounding analysis was performed by eliminating failure of air-operated valves to open in the PSA model [gates M41-002, M41-002-SBO, and M41-002X were set to zero], which resulted in an internal and external benefit (with uncertainty) of approximately $93,240. This analysis case was used to model the benefit of Phase II SAMAs 46 and 47.Case 34:ISLOCA This analysis case was used to evaluate the change in plant risk from reducing the probability of an ISLOCA by increasing the frequency of valve leak testing or improving ISLOCA identification or coping. A bounding analysis was performed by setting the ISLOCA initiators to zero in the PSA model [initiators %VPCIC, %VLPCS, and %VSDC were set to zero], which resulted in an internal and external benefit (with uncertainty) of approximately $231. This analysis case was used to model the benefit of Phase II SAMAs 48, 50, and 51.Case 35:MSIV DesignThis analysis case was used to evaluate the change in plant risk from improving MSIV design to decrease the likelihood of containment bypass scenarios. A bounding analysis was performed by eliminating failure of the MSIVs to close or remain closed in the PSA model [gates DL-MSIV, IS-MSIV, and IS-MSIV-INIT were removed from the model], which resulted in an internal and external benefit (with uncertainty) of approximately $30,093. This analysis case was used to model the benefit of Phase II SAMA 49.Case 36:Standby Liquid Control (SLC) SystemThis analysis case was used to evaluate the change in plant risk from increasing boron concentration in the SLC system. A bounding analysis was performed by eliminating the contribution due to failure to initiate SLC and failures of alternate boron injection in the PSA model [gate SLC was removed from the model and basic event ABI was set to zero], which resulted in an internal and external benefit (with uncertainty) of approximately $31,849. This analysis case was used to model the benefit of Phase II SAMA 52.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-12Case 37: SRV ReseatThis analysis of case was used to evaluate the change in plant risk from installing more reliable SRVs. A bounding analysis was performed by eliminating the initiator for the SRVs inadvertently being open and the basic events for stuck open SR Vs in the PSA model [initiator %T3C, basic events P1 and P2 were set to zero], which resulted in an internal and external benefit (with uncertainty) of approximately $87,324. This analysis case was used to model the benefit of Phase II SAMA 53.Case 38:Add Fire SuppressionThis analysis case was used to evaluate the change in plant risk from adding automatic fire suppression systems to the dominant fire zones. The dominant fire zones reported in the IPEEE are the control room and control building switchgear rooms. The control room has Halon suppression in the control room floor sections. Many of the switchgear rooms have automatic CO 2 suppression systems. The Div I switchgear room in the control building that is a large contributor in the IPEEE is zone OC202 in compartment CC202, which has a partial automatic sprinkler system. For the main control, an automatic suppression system would not provide a significant safety benefit. The sensing devices used for fires include both fuse elements that melt given high temperature and smoke detectors. These types of actuation devices would only actuate after the fire has progressed to a point that would cause evacuation of the control room. Even if the auto suppression system actuated prior to evacuation, the consequences of actuation would require evacuation. Additional Halon or CO 2 systems would asphyxiate any personnel remaining in the main control room and water would damage the control equipment. Given that the main control room fire risk is dominated by failure to shut down the reactor from outside the control room, extremely limited benefit is judged to exist for auto suppression systems in the main control room.Thus, this SAMA evaluates improving the reliability and effectiveness of the suppression systems in the switchgear rooms. A bounding analysis was performed as described below, which resulted in an internal and external benefit (with uncertainty) of approximately $102,345. This analysis case was used to model the benefit of Phase II SAMA 54.This analysis case (Adding automatic fire suppression systems to the critical switchgear rooms) is an external events SAMA, which would not mitigate internal event risk. Many of the switchgear rooms have automatic CO 2 suppression systems. The Div I switchgear room in the control building that is a large contributor in the IPEEE is zone OC202, which has a partial automatic sprinkler system. This SAMA would improve the reliability and effectiveness of those systems. A bounding analysis was performed by assuming the SAMA would eliminate the contribution to fire CDF from fires in critical switchgear room OC202. Since the total fire CDF is 2.74E-05/yr

[TableE.1-10] and the critical switchgear room fire CDF is 9.37E-07/yr, fires in the critical switchgear rooms contribute 3.42% of the total fire CDF.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-13The internal events model cannot be used to assess the benefit from this external event SAMA. However, the consequences resulting from fire-induced core damage and internal event-induced core damage would be comparable. Since we have already estimated the maximum benefit from removing all internal event risk, the maximum benefit of removing all fire risk was estimated by reducing the maximum internal event benefit by the ratio of the total fire CDF to the internal event CDF. Since this SAMA analysis case would eliminate 3.42% of the total fire risk, the benefit for this SAMA analysis case was estimated to be 3.42% of the total fire benefit as shown below.

Given,Maximum internal benefit is $74,673 [Table4.21-1

]Total fire CDF = 2.74E-05/rx-yr [TableE.1-10

]Internal events CDF = 2.05E-06/rx-yrMaximum fire benefit = Maximum internal benefit x Total fire CDF/Internal events CDF Maximum fire benefit = $74,673 x (2.74E-05/2.05E-06)= $997,559SAMA case 38 benefit = 3.42% x (Maximum fire benefit) = 0.0342 x $997,559 SAMA case 38 benefit = $34,115Applying the uncertainty factor of 3,SAMA case 38 benefit with uncertainty = $34,115 x 3 = $102,345Case 39:Reduce Risk from Fires that Require Control Room EvacuationThe alternate shutdown system (ASDS) panel is designed to use division 1 safety and support systems to safely shutdown the plant. This analysis case was used to evaluate the change in plant risk from upgrading the ASDS panel to include additional system controls for the other division. A bounding analysis was performed as described below, which resulted in an internal and external benefit (with uncertainty) of approximately $420,521. This analysis case was used to model the benefit of Phase II SAMA 55.This SAMA analysis case is an external events SAMA, which would not mitigate internal event risk. A bounding analysis was performed by assuming the SAMA would eliminate the contribution to fire CDF from fires in the control room. Since the total fire CDF is 2.74E-05/yr and the control room fire CDF is 3.85E 06/yr, fires in the control room contribute 14.05% of the total fire CDF. The internal events model cannot be used to assess the benefit from this external event SAMA. However, the consequences resulting from fire-induced core damage and internal event-induced core damage would be comparable. Since we have already estimated the maximum benefit from removing all internal event risk, the maximum benefit of removing all fire risk can be estimated by reducing the maximum internal event benefit by the ratio of the total fire CDF to the internal event CDF. Since this SAMA analysis case would eliminate 14.05% of the total fire risk, the benefit for this SAMA analysis case was estimated to be 14.05% of the total fire benefit as shown below.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-14 Given,Maximum internal benefit is $74,673 [Table4.21-1

] Total fire CDF = 2.74E-05/rx-yr [TableE.1-10

]Internal events CDF = 2.05E-06/rx-yrMaximum fire benefit = Maximum internal benefit x Total fire CDF/Internal events CDFMaximum fire benefit = $74,673 x (2.74E-05/2.05E-06) = $997,599SAMA case 39 benefit = 14.05% x (Maximum fire benefit) = 0.1405 x $997,599 SAMA case 39 benefit = $140,174Applying the uncertainty factor of 3,SAMA case 39 benefit with uncertainty = $140174 x 3 = $420,521Case 40: Large Break LOCAThis analysis case was used to evaluate the change in plant risk from installing a digital large break LOCA protection system. A bounding analysis was performed by setting the large LOCA initiator to zero in the PSA model [initiator %A was set to zero], which resulted in an internal and external benefit (with uncertainty) of approximately $948,372. This analysis case was used to model the benefit of Phase II SAMA 56.Case 41:Trip/Shutdown RiskThis analysis case was used to evaluate the change in plant risk from implementing Generation Risk Assessment (trip and shutdown risk modeling) in plant activities. It is assumed that this would reduce the frequency of plant trips and shutdowns. A bounding analysis was performed by reducing all initiating event frequencies except pipe breaks, floods, and LOSP by 10% [the following initiating events were reduced: %T2, %T2M, %T3A, %T3B, %T3C, %TAC1, %TAC2, %TBCW, %TCCW, %TCRD, %TDC1, %TDC2, %TIA, %TPSW, %TST11, and %TST21], which resulted in an internal and external benefit (with uncertainty) of approximately $187,117. This analysis case was used to model the benefit of Phase II SAMA 57.Case 42:Increase Availability of SSW Pump House Ventilation SystemThis analysis case was used to evaluate the change in plant risk from increasing the training emphasis and providing additional control room indication on the operational status of the SSW pump house ventilation system. This will allow operators to manually open the pump house dampers, which can provide adequate ventilation such that pump failures would not occur. A bounding analysis was performed by eliminating failure of SSW Pump House Ventilation in the PSA model [the following gates were removed from the model: HVC-1000X, HVC-1000XP, HVC-1000XZ, HVC-1000-HPCS, HVC-1000X-HPCS, HVC-1000X-ONSP, HVC-1000X-SBO, and HVC-2000X], which resulted in an internal and external benefit (with uncertainty) of approximately $45,212. This analysis case was used to model the benefit of Phase II SAMA 58.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-15Case 43:Increase Recovery Time of ECCS upon Loss of SSWThis analysis case was used to evaluate the change in plant risk from upgrading procedures and increasing operator training for alternating operation of the low pressure ECCS pumps (LPCI and LPCS) for loss of SSW scenarios. A bounding analysis was performed by eliminating failure of the SSW to the LPCS room cooler in the PSA model [gate P41-LPCS was removed from the model], which resulted in an internal and external benefit (with uncertainty) of approximately $121,357. This analysis case was used to model the benefit of Phase II SAMA 59.Case 44:Additional Containment Heat RemovalThis analysis of case was used to evaluate the change in plant risk from installing an additional method of removing heat from the containment. A bounding analysis was performed by eliminating failure of suppression pool cooling and containment spray systems in the PSA model [the following gates were removed from the model: RH--SY-SPCSYS-F-, E12-199, E12-199X, E12-199XX, E12-199X-SBO, E12-199Y, E12-199Y-SBO, E12-199-SBO, E12-199-CSS, E12-600, E12-600X, E12-600XX, E12-600X-SBO, E12-600Y, E12-600Y-SBO, and E12-600-SBO],

which resulted in an internal and external benefit (with uncertainty) of approximately $894,362.

This analysis case was used to model the benefit of Phase II SAMA 60.Case 45:Improve RHR Heat Exchanger AvailabilityThis SAMA analysis case was used to evaluate the change in plant risk from adding a bypass around the RHR HX inlet and outlet valves. A bounding analysis was performed by eliminating failure of RHR HX Cooler inlet and outlet valves in the PSA model [the following basic events were set to zero: P41-CC-MVF014A-L, P41-CC-MVF014B-L, P41-CC-MVF068A-L, P41-CC-MVF068B-L, P41-CF-MVF14AB-L, and P41-CF-MVF68AB-L], which resulted in an internal and external benefit (with uncertainty) of approximately $124,019. This analysis case was used to model the benefit of Phase II SAMA 62.Case 46:Improve RCIC Lube Oil CoolingThis analysis case was used to evaluate the change in plant risk from adding a redundant RCIC lube oil cooling path. A bounding analysis was performed by eliminating the failure to cool RCIC lube oil in the PSA model [gate E51-043-G was set to zero], which resulted in an internal and external benefit (with uncertainty) of approximately $92,683. This analysis case was used to model the benefit of Phase II SAMA 63.E.2.4Sensitivity AnalysesTwo sensitivity analyses were conducted to gauge the impact of assumptions upon the analysis. The benefits estimated for each of these sensitivities are presented in TableE.2-3

.A description of each sensitivity case follows.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-16Sensitivity Case 1: Years Remaining Until End of Plant LifeThe purpose of this sensitivity case was to investigate the sensitivity of assuming a 33-year period for remaining plant life (i.e., thirteen years on the original plant license plus the 20-year license renewal period), rather than the 20-year license renewal period used in the base case. Changing this assumption does not cause additional SAMAs to be cost-beneficial. Sensitivity Case 2: Conservative Discount RateThe purpose of this sensitivity case was to investigate the sensitivity of each analysis case to the discount rate. The discount rate of 7.0% used in the base case analyses is conservative relative to corporate practices. Nonetheless, a lower discount rate of 3.0% was assumed in this case to investigate the impact on each analysis case. Changing this assumption does not cause additional SAMAs to be cost-beneficial.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-17E.2.5ReferencesE.2-1Nuclear Energy Institute (NEI), NEI 05-01, Severe Accident Mitigation Alternatives (SAMA) Analysis Guidance Document, November 2005, Revision A.E.2-2U.S. Nuclear Regulatory Commission (USNRC), NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding James A. FitzPatrick Nuclear Power Plant (NUREG-1437, Supplement 31) Final Report, January 2008.E.2-3USNRC, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Vermont Yankee Nuclear Power Station (NUREG-1437, Supplement 30) Final Report, August 2007.E.2-4USNRC, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Pilgrim Nuclear Power Station (NUREG-1437, Supplement

29) Final Report, July 2007.E.2-5USNRC, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Oyster Creek Nuclear Generating Station (NUREG-1437, Supplement 28) Final Report, January 2007.E.2-6USNRC, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Monticello Nuclear Generating Plant (NUREG-1437, Supplement 26) Final Report, August 2006.E.2-7USNRC, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Brunswick Steam Electric Plant, Units 1 and 2 (NUREG-1437, Supplement 25) Final Report, April 2006.E.2-8USNRC, NUREG-1742 Perspectives Gained From the Individual Plant Examination of External Events (IPEEE) Program, Volumes 1 & 2, Final Report April 2002.E.2-9USNRC, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Cooper Nuclear Station (NURE G-1437, Supplement 41) Final Report , July 2010.E.2-10USNRC, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Susquehanna Steam Electric Station, Units 1 and 2 (NUREG-1437, Supplement 35) Final Report, March 2009.E.2-11USNRC, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Duane Arnold Energy Center (NUREG-1437, Supplement 42) Final Report, October 2010.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-18E.2-12USNRC, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Hope Creek Generating Station and Salem Nuclear Generating Station, Units 1 and 2 (NUREG-1437, Supplement 45) Final Report, March 2011.E.2-13GGNS94-0054 IPEEE "Internal Plant Examination of External Events Summary Report," Revision 1.E.2-14GGNS95-00041 IPEEE "Internal Plant Examination of External Events Fire," Revision 0.E.2-15GGNS94-0053 IPEEE "Internal Plant Examination of External Events Seismic Margins," Revision 0.E.2-16GGNS94-0051IPEEE "Internal Plant Examination of External Events Fire Modeling," Revision 1.E.2-17GGNS93-0048 IPEEE "Internal Plant Examination of External Events High Wind and Tornado Assessment," Revision 0.E.2-18Grand Gulf Nuclear Station Individual Plant Examination Summary Report (IPE), December 1992.E.2-19Energy Northwest, License Renewal Application, Appendix E, "Applicant's Environmental Report, Operating License Renewal Stage, Columbia Generating Station, Energy Northwest," January 2010.E.2-20USNRC, NUREG/BR-0184, Regulatory Analysis Technical Evaluation Handbook , January 1997.E.2-21CALC-OC-N1000-10003, "Cost-Benefit Analysis of Severe Accident Mitigation Alternatives," Revision 0.

Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-19Table E.2-1Phase I SAMAs Related to IPE and IPEEE InsightsPhase I SAMA ID NumberSAMA TitleResult of Potential EnhancementScreening ResultsSAMA DispositionCredited in PSA Model226The Loss of Offsite Power Off-Normal Event Procedure will be revised to allow for the Level 2 signal to be bypassed in the event that the Division 3 diesel generator must be cross-tied to Divisions 1 or

2.Increased availability of on-site AC power leading to increased availability of

ECCS injection.#3 - Already installedThe Loss of AC Power Off-Normal Event Procedure has been revised to allow for the level 2 signal to be bypassed in the event that the Division 3 diesel generator must be cross-tied to divisions 1 or 2.Yes Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-20227Improve secondary containment isolation to allow the capability of bypassing the isolation signals and re-opening the valves.Improved availability of PSW and Instrument Air such that the main condenser, condensate, and feedwater systems would not be lost. CRD would also not be degraded due to a loss of the preferred cooling source of the component cooling water (CCW) heat exchangers.#3 - Already installedThe PSW isolation valves in the Auxiliary Building penetrations (P44-F1l6, P44-FI17, P44-FII8, P44-FII9, P44-FI20, P44-FI2I, P44-FI22 and P44-FI23) can be reopened by manual override after a LOCA to reestablish PSW cooling to the CCW heat exchangers, computer room coolers, plant chillers, steam tunnel coolers, and drywell coolers. This should be done only if offsite power is available and after it has been determined that the release of radioactive fission products will not

result.05-S-01-EP-1 contains guidance to restore instrument air to containment loads by defeating containment isolation interlocks and opening the valves.YesTable E.2-1 (Continued)Phase I SAMAs Related to IPE and IPEEE InsightsPhase I SAMA ID NumberSAMA TitleResult of Potential EnhancementScreening ResultsSAMA DispositionCredited in PSA Model Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-21228Implement procedural changes to allow for bypass of the RCIC turbine trip due to main steam tunnel (MST) high temperature when PSW is unavailable and no steam line break has occurred.Increased RCIC availability when main steam tunnel high temperature exists.#1 - N / AProvided there is no leak in the main steam tunnel, failure of main steam cooling will not result in a MST temperature of 185°F or greater. Therefore, it will not result in an initiation of the MST high temperature isolation logic.

No229Increase the training emphasis and provide additional control room indication on the operational status of the SSW pump house ventilation system. This will allow operators to manually open the pump house dampers, which can provide adequate ventilation such that pump failures would not occur.Increased availability of the SSW pump house ventilation system.Retain (Phase II

SAMA 58)In accordance with GDC 13, damper status is indicated in the main control room. In addition, there is a high temperature alarm in the main control room.Alarm 04-1-02-1H13-P870 provides an alarm, but the actions could be expanded to accomplish a more robust mitigation of this condition.

No230Increase operator training for alternate operation of the low pressure ECCS pumps (LPCI and LPCS) for loss of SSW scenarios.Increased time available for recovery actions for low pressure ECCS when a loss of SSW occurs.Retain (Phase II

SAMA 59)No specific operator training is in place to address this condition.

NoTable E.2-1 (Continued)Phase I SAMAs Related to IPE and IPEEE InsightsPhase I SAMA ID NumberSAMA TitleResult of Potential EnhancementScreening ResultsSAMA DispositionCredited in PSA Model Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-22231Revise the containment flooding portion of the Emergency Procedures to remove or modify the step requiring MSIV venting.Limit one of the major contributors to the source term released.#3 - Already installedGGNS contributed this IPE insight to the BWR Owners Group Severe Accident Subcommittee. GGNS has already implemented the current SAGs on RPV venting.Yes232Install a backup power supply to the hydrogen igniters.Hydrogen igniter operability during station blackout.#3 - Already InstalledGGNS has two hydrogen recombiners, each powered from a different division. They are backed up by hydrogen igniters and a drywell purge system. Also, GGNS has a portable generator used to supply temporary power to one division of hydrogen igniters.

No233Install an additional method of removing heat from the containment.Increased decay heat removal capabilityRetain (Phase II

SAMA 60)GGNS utilizes the containment spray and RHR suppression pool cooling for post-accident containment heat removal.

Containment venting is also available to ensure pressure stays below design limits should the other systems fail to reduce containment

pressure.NoTable E.2-1 (Continued)Phase I SAMAs Related to IPE and IPEEE InsightsPhase I SAMA ID NumberSAMA TitleResult of Potential EnhancementScreening ResultsSAMA DispositionCredited in PSA Model Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-23234Install a backup water supply and pumping capability that is independent of normal and emergency AC power.Alternate water supply for containment spray/vessel injectionRetain (Phase II

SAMA 61)GGNS has a high pressure core spray system, which is powered from an independent (Division 3) power supply; however, a backup supply will be investigated per the IPE recommendations.

No235Extend the battery depletion time for the relief valves.Enhanced reactor pressure vessel depressurization system

reliability

  1. 2 - Similar item is addressed under other proposed SAMAsADS and Non-ADS relief valves are all dependent on DC power and instrument air. Extended DC power to the relief valves will allow longer operation during a loss of DC battery chargers.Similar to Phase II SAMAs 1, 3, and 27.No236Implement the latest revision of the BWR Owners Group emergency procedure guidelines (EPGs).Improved likelihood of success of operator actions taken in response to abnormal conditions.#3 - Already Installed GGNS currently utilizes Revision 2 of the BWROG EPGs. Yes237Increase maintenance on drainage structures. Maintenance should include cleaning of culverts, concrete repair, and removal of vegetation/debris which could obstruct flow.Prevent deterioration of site conditions.#3 - Already installedGGNS has increased the maintenance on drainage

structures.

NoTable E.2-1 (Continued)Phase I SAMAs Related to IPE and IPEEE InsightsPhase I SAMA ID NumberSAMA TitleResult of Potential EnhancementScreening ResultsSAMA DispositionCredited in PSA Model Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-24238Plant procedures currently require plant staff to insure that plant doors are closed during severe weather and in the event of plant flooding (Implicitly including former Unit 2 doors). Revise procedures to explicitly include at-grade former Unit 2 doors.Reduce leakage from flooding through an open door.#3 - Already installedGGNS has revised the plant flood mitigation procedure.

No239Revise procedures to periodically inspect roof drains and overflows to ensure they are not blocked.Reduce the consequences of a flood.#3 - Already installedGGNS has created an inspection procedure for roof drains, roof drainage system, and roof overflows.

NoTable E.2-1 (Continued)Phase I SAMAs Related to IPE and IPEEE InsightsPhase I SAMA ID NumberSAMA TitleResult of Potential EnhancementScreening ResultsSAMA DispositionCredited in PSA Model Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-25240Remove the wooden foot bridge crossing the northwest ditch near its upstream end.Improve site drainage/external flood protection.#1 - N/AThe IPEEE showed the risk from external flooding at GGNS is minor.

Thus this potential modification is assumed not to be cost beneficial, which follows the same assumption in the NRC safety evaluation report. In May 2011, NRC Inspectors verified that the plant grade is 132.5 feet above mean sea level and that the maximum expected flood height from the Mississippi River is about 103 feet above mean sea level.

Therefore, floodwaters from the Mississippi River are not expected to impact the plant.

NoTable E.2-1 (Continued)Phase I SAMAs Related to IPE and IPEEE InsightsPhase I SAMA ID NumberSAMA TitleResult of Potential EnhancementScreening ResultsSAMA DispositionCredited in PSA Model Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-26241Remove the 15" corrugated metal pipe located in the small auxiliary ditch parallel to the northwest ditch (at the same approximate location as the duct bank crossing the northwest ditch). Re-grade the area to provide a gradual transition between the yard upstream and the auxiliary ditch.Improve site drainage/external flood protection.#1 - N/AThe IPEEE showed the risk from external flooding at GGNS is minor.

Thus this potential modification is assumed not to be cost beneficial, which follows the same assumption in the NRC safety evaluation report.In May 2011, NRC Inspectors verified that the plant grade is 132.5 feet above mean sea level and that the maximum expected flood height from the Mississippi River is about 103 feet above mean sea level.

Therefore, floodwaters from the Mississippi River are not expected to impact the plant.

NoTable E.2-1 (Continued)Phase I SAMAs Related to IPE and IPEEE InsightsPhase I SAMA ID NumberSAMA TitleResult of Potential EnhancementScreening ResultsSAMA DispositionCredited in PSA Model Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-27242Re-hang the security fence gates west of the control building to insure that approximately 5" of gap exists between the gate and the road.Improve site drainage/external flood protection.#1 - N/AThe IPEEE showed the risk from external flooding at GGNS is minor.

Thus this potential modification is assumed not to be cost beneficial, which follows the same assumption in the NRC safety evaluation report.In May 2011, NRC Inspectors verified that the plant grade is 132.5 feet above mean sea level and that the maximum expected flood height from the Mississippi River is about 103 feet above mean sea level.

Therefore, floodwaters from the Mississippi River are not expected to impact the plant.

NoTable E.2-1 (Continued)Phase I SAMAs Related to IPE and IPEEE InsightsPhase I SAMA ID NumberSAMA TitleResult of Potential EnhancementScreening ResultsSAMA DispositionCredited in PSA Model Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-28243Grade down and remove the access road, the raised berm parallel to the access road, and curbs adjacent to the access road as necessary where they cross Culvert No.1, such that elevations above the culvert do not exceed 132.7 ft. MSL.Improve site drainage/external flood protection.#1 - N/AThe IPEEE showed the risk from external flooding at GGNS is minor.

Thus this potential modification is assumed not to be cost beneficial, which follows the same assumption in the NRC safety evaluation report.In May 2011, NRC Inspectors verified that the plant grade is 132.5 feet above mean sea level and that the maximum expected flood height from the Mississippi River is about 103 feet above mean sea level.

Therefore, floodwaters from the Mississippi River are not expected to impact the plant.

NoTable E.2-1 (Continued)Phase I SAMAs Related to IPE and IPEEE InsightsPhase I SAMA ID NumberSAMA TitleResult of Potential EnhancementScreening ResultsSAMA DispositionCredited in PSA Model Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-29244Replace the C8x11.5 channel forming the flood barrier across the SSW A equipment hatch opening with another member having a minimum depth of approximately 13".Improve site drainage/external flood protection.#1 - N/AThe IPEEE showed the risk from external flooding at GGNS is minor.

Thus this potential modification is assumed not to be cost beneficial, which follows the same assumption in the NRC safety evaluation report.In May 2011, NRC Inspectors verified that the plant grade is 132.5 feet above mean sea level and that the maximum expected flood height from the Mississippi River is about 103 feet above mean sea level.

Therefore, floodwaters from the Mississippi River are not expected to impact the plant.

No245Modify the piping systems to account for the grouted condition for the penetration of the standby service water (SSW) piping in the

control building.

Reduce vulnerability to a seismic event.#3 - Already installedThe grout was removed and the pipe support at the penetration was modified to coincide with the design basis piping analysis assumption.

NoTable E.2-1 (Continued)Phase I SAMAs Related to IPE and IPEEE InsightsPhase I SAMA ID NumberSAMA TitleResult of Potential EnhancementScreening ResultsSAMA DispositionCredited in PSA Model Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-30Table E.2-2Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion1. DC Power Eliminates all SBO cutsets13.6%16.5%13.6% $115,656 $346,9681 - Provide additional DC battery capacityCNS estimate.$500,000Not cost effective2 - Replace lead-acid batteries with fuel

cellsCNS estimate.$1,000,000Not cost effective11 - Portable generator for direct current (DC) power:

This SAMA involves the use of a portable generator to supply DC power to the battery chargers during a station blackout.CNS had different cost estimates for the portable generator to supply the charger and the portable generator to supply a panel because they had an existing generator big enough to supply the charger, but not big enough to supply a panel.(cont. below)$714,000Not cost effective Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-31(cont.)Since GGNS does not have an existing generator that can be used for either purpose, the CNS estimate for a new generator is appropriate. Thus, GGNS SAMA 11 cost

estimate should be the same as GGNS SAMA 12 cost

estimate.12 - Portable generator for direct current (DC) power:

This SAMA involves the use of a portable generator to supply DC power to the individual panels during a station blackout.CNS estimate.$714,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-3215 - Use DC generators to provide power to operate the switchyard power control breakers while a 480-V AC generator could supply the air compressors for breaker support.GGNS SAMA 11 and SAMA 12 estimate that one generator would cost

~$714,000. This

SAMA recommends addition of at least two generators.

Thus, GGNS SAMA 15 cost estimate should be at least double that for SAMA 11 or 12.$1,428,000Not cost effective2. Improve Charger ReliabilityFailure of chargers contribution to zero.1.4%2.2%2.3% $13,598 $40,793 3 - Add battery charger to existing DC systemCNS estimate.$90,000Not cost effective13 - Proceduralize battery charger high-voltage shutdown circuit inhibitCNS estimate.$50,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-333. Add DC System Cross-ties Eliminate failure of DC power gates.7.6%11.6%11.8% $73,056 $219,169 4 - Provide DC bus cross-tiesCNS estimate.$300,000Not cost effective4. Increase Availability of On-Site AC Power Eliminated failure of DG11, DG12, and DG13 to their AC

Busses17.5%21.2%18.5% $149,396 $448,1895 - Provide an additional diesel generatorCNS estimate.$20,000,000Not cost effective8 - Install a gas turbine generator with tornado protectionCNS estimate.$2,000,000Not cost effective5. Improve AC PowerEliminated the loss of the 4.16-kV buses20.4%25.6%23.2% $177,524 $532,5716 - Improve 4.16-kV bus cross-tie abilityCNS estimate.$656,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-3417 - Provide alternate feeds to essential loads directly from an alternate emergency busModification of the AC system to allow alignment of alternate feeds to the 4kV loads is greater in scope than an AC crosstie modification.

SAMA 6, Improve 4.16-kV bus cross-tie ability, is estimated to cost $656,000. Thus, this is a lower bound estimate for SAMA 17.$656,000Not cost effective6. Reduce Loss of Off-Site Power

During Severe Weather Eliminate the weather centered loss of off-site power initiating event.3.1%3.7%3.1% $26,087 $78,2617 - Install an additional, buried off-site power source. CNS estimate.$2,485,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-357. Provide Backup EDG Cooling Eliminated failure of SW cooling to the

EDGs1.9%2.5%1.9% $16,515 $49,5459 - Use fire water system as backup source for diesel

cooling Hardware modification range

estimate.$100,000Not cost effective10 - Add new backup source of diesel

coolingCNS estimate.$2,000,000Not cost effective8. Increase EDG Reliability Eliminated failure of EDGs to run3.3%4.6%4.5% $30,348 $91,04414 - Provide a portable EDG fuel oil transfer pumpCNS estimate.$100,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-369. Improve DG ReliabilityEliminated the common cause

failure (CCF) contribution of failure to start EDGs0.3%0.3%0.2% $2,181 $6,54216 - Provide a diverse swing diesel generator air start compressor Hardware modification range

estimate.$100,000Not cost effective10. Reduce Plant-Centered Loss of Off-Site PowerRemoved the contribution of plant-and switchyard-centered events9.1%10.7%8.9% $76,556 $229,668 18 - Protect transformers from failureCNS estimate.$780,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-3711. Redundant Power to Torus Hard Pipe Vent (THPV)

Valves Eliminated failure of power to containment vents1.1%1.8%1.8% $10,766 $32,29719 - Provide redundant power to direct torus hard pipe vent valves to

improve the reliability of the direct torus vent valves and enhance the containment heat removal capability.CNS estimate.$714,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-3812. High Pressure Injection System Eliminated failure of the HPCS77.8%61.8%60.2% $594,912 $1,784,736 20 - Install an independent active or passive high pressure injection systemRecent BWR cost estimates for this SAMA are ~$2M at Duane Arnold, ~$4M at Susquehanna,

~$5M at Vermont Yankee, and ~$29M at Columbia.SAMA 24, Add a diverse low pressure injection system, is estimated to cost

$8,800,000. Since a high pressure system would cost at least as much as a low pressure system, this estimate is appropriate.$8,800,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-3961 - Install a backup water supply and pumping capability that is independent of normal and emergency AC powerPlant-specific cost estimate.$6,409,949Not cost effective 13. Extend RCIC Operation Eliminated failure of trip due to pressure1.0%1.6%1.7% $10,031 $30,09321 - Raise HPCI/RCIC backpressure trip set points [HPCI backpressure trip setpoint has already been raised. This SAMA will evaluate raising the RCIC backpressure trip set point]. CNS estimate.$200,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-4014. Improve ADS System Eliminated failure of ADS valves45.9%16.3%16.0% $299,106 $897,317 22 - Modify automatic depressurization system components

to improve reliability

[This SAMA will add larger accumulators thus increasing

reliability during

SBOs].Plant-specific cost estimate.$1,176,850Not cost effective15. Improve ADS Signals Eliminated failure of the SRV failing to open20.8%5.3%4.8% $129,383 $388,15023 - Add signals to open safety relief valves automatically in an MSIV closure

transient. CNS estimate.$1,500,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-4116. Low Pressure Injection System Eliminated failure of the LPCI and LPCS22.9%39.5%38.3% $229,965 $689,89624 - Add a diverse low pressure injection system.CNS estimate.$8,800,000Not cost effective17. ECCS Low Pressure InterlockEliminated ECCS permissives and

interlock failure1.0%1.6%1.7% $10,031 $30,09325 - Install a bypass switch to allow operators to bypass the low reactor pressure interlock circuitry that inhibits opening the LPCI or core spray injection

valves following sensor or logic failures that prevent

all low pressure injection valves from opening.CNS estimate.$1,000,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-4218. RHR Heat Exchangers Eliminated failure of SSW to provide cooling to the RHR heat exchangers18.5%37.4%39.9% $205,223 $615,66926 - Implement modifications to allow manual alignment of the fire water system to RHR heat exchangers.Pilgrim estimate.$1,950,000Not cost effective19. Emergency Service Water System Reliability Eliminated failure of service water pumps3.6%6.7%7.0% $37,903 $113,708 27 - Add a service water pump to increase availability of

cooling waterCNS estimate.$5,900,000Not cost effective20. Main Feedwater System Reliability Eliminated failure to inject from feedwater19.3%20.5%20.6% $162,050 $486,14928 - Add a motor-driven feed water pumpCNS estimate.$1,650,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-4321. Increase Availability of Room Cooling Eliminated failure of room cooling to LPCS, HPCS, SSW and safeguard switchgear battery rooms22.9%17.8%18.2% $175,400 $526,20029 - Provide a redundant train or means of ventilationCNS estimate.$2,202,725Not cost effective22. Increase Availability of the DG System Through HVAC Improvements Eliminated failure of diesel generator rooms HVAC9.2%10.6%8.5% $75,988 $227,96330 - Add a diesel building high temperature alarm or redundant louver and thermostat.CNS estimate.$1,304,700Not cost effective32 - Diverse EDG HVAC logicCost for Phase II SAMAs 4 and 31 is used because the modifications are similar in scope.$300,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-4433 - Install additional fan and louver pair for EDG heating, ventilation, and air conditioningCNS estimate.$6,000,000Not cost effective23. Increased reliability of HPCI and RCIC room cooling Eliminated failure of power to the HPCS pump room cooler.

(RCIC pump continued operation is not dependent on room cooling.)1.0%1.6%1.7% $10,031 $30,09331 - Create ability to switch HPCI and RCIC room fan power supply to DC in an SBO event.CNS estimate. Similar to SAMA 4, provide DC bus cross-ties.$300,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-4524. Increase Reliability of

Instrument Air Eliminated failure of the instrument air14.9%20.2%21.3% $137,842 $413,527 34 - Modify procedure/hardware to provide ability to align diesel power to more air compressorsCNS estimate. More than just procedure.$1,200,000Not cost effective35 - Replace service and instrument air compressors with more reliable compressors which have self-contained air cooling by shaft-driven fansCNS estimate.$1,394,598Not cost effective25. Backup Nitrogen to SRVEliminated operator failure to install air bottles5.5%3.7%3.8% $40,614 $121,84136 - Install nitrogen bottles as backup gas supply for safety relief

valves.Plant-specific cost estimate.$1,722,706Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-4626. Improve Availability of SRVs and MSIVs Eliminated failure of non-ADS SRVs46.1%16.4%16.1% $300,631 $901,89337 - Improve SRV and MSIV pneumatic components.CNS estimate.$1,500,000Not cost effective27. Improve Suppression Pool Cooling Eliminated the failure of flow to the RHR heat exchangers18.5%37.4%39.9% $205,223 $615,669 38 - Install an independent method of suppression pool cooling.CNS estimate.$5,800,000Not cost effective28. Increase Availability of Containment Heat

Removal Eliminated failure of cooled flow from RHR pump A and B26.6%51.6%54.7% $288,437 $865,31239 - Procedural change to cross-tie open cycle cooling system to enhance containment spray systemProcedural range estimate.$25,000RetainTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-4741 - Use the fire water system as a backup source for the drywell spray systemSimilar to Phase II SAMA 26, implement modifications to allow manual alignment of the fire water system to RHR heat exchangers.$1,950,000Not cost effective29. Decay Heat Removal Capability

- Drywell Spray Eliminated failure of RHR spray26.6%51.6%54.7% $288,550 $865,64940 - Install a passive drywell spray system to provide redundant drywell spray method.CNS estimate.$5,800,000Not cost effective30. Increase Availability of the

CST Eliminated failure of HPCS and RCIC suction11.3%16.8%17.4% $107,899 $323,69642 - Enhance procedures to refill

CST from demineralized water or service water system.Procedure with engineering and training range

estimate.$200,000RetainTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-4831. Filtered Vent to Increase Heat Removal Capacity for Non-ATWS EventsReduced the baseline accident progression source terms by a factor of 20.0%26.4%34.3% $80,920 $242,759 43 - Install a filtered containment vent to provide fission product scrubbingCNS estimate.$1,500,000Not cost effective32. Reduce Hydrogen Ignition Eliminated failure of hydrogen igniters15.9%20.7%20.2% $142,455 $427,365 44 - Provide post-accident containment inerting capability.Plant-specific cost estimate.$2,665,123Not cost effective45 - Install a passive hydrogen control system.Monticello (SAMA 10) estimated that this modification would cost

~$760,000.$760,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-4933. Controlled Containment Venting Eliminated failure of air-operated valves

to open2.9%5.4%5.8% $31,080 $93,24046 - Provide passive overpressure relief by changing the containment vent valves to fail open and improving the strength of the rupture diskCNS estimate.$1,000,000Not cost effective47 - Enable manual operation of all containment vent valves via local controlsOyster Creek (SAMA 84) estimated that it would cost $150,000 to add handwheels in the reactor building to open AOVs in the current vent path.$150,000Not cost effective34. ISLOCARemoved all ISLOCA initiators< 0.1%< 0.1%< 0.1% $77 $23148 - Increase frequency of valve leak testing to reduce ISLOCA frequencyCNS estimate.$100,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-5050 - Revise EOPs to improve ISLOCA identificationCNS estimate.$50,000Not cost effective51 - Improve operator training on ISLOCA copingCNS estimate.$112,000Not cost effective35. MSIV Design Eliminated failure of the MSIVs to close or remain closed1.0%1.6%1.7% $10,031 $30,09349 - Improve MSIV design to decrease the likelihood of containment bypass scenarios.CNS estimate.$1,000,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-5136. SLC System Eliminated failure to initiate SLC and failures of alternate boron injection (ABI)1.1%1.7%1.7% $10,616 $31,84952 - Increase boron concentration in the SLC system

[Reduced time required to achieve shutdown provides

increased margin in the accident timeline for successful initiation of SLC]CNS estimate.$50,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-5237. SRV ReseatEliminated the initiator for SRVs inadvertently being open and basic events for stuck open SRVs3.1%4.3%4.5% $29,108 $87,32453 - Increase safety relief valve (SRV)

reseat reliability to address the risk associated with dilution of boron caused by the failure of the SRVs to reseat after standby liquid control (SLC) injectionCNS estimate.$2,200,000Not cost effective38. Add Fire Suppression 1Eliminated fire CDF from the critical switchgear rooms.n/an/an/a $34,115 $102,34554 - Add automatic fire suppression systems to the dominant fire zonesCNS estimate.$375,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-5339. Reduce Risk from Fires that

Require Control Room Evacuation (1)Eliminate fire CDF from the main control room.n/an/an/a $140,174 $420,52155 - Upgrade the ASDS panel to include additional system controls for opposite division.CNS estimate.$786,991Not cost effective40. Large Break LOCAEliminated Large Break LOCA7.1%16.5%17.5% $316,124 $948,37256 - Provide digital large break LOCA protection to identify symptoms/precursors of a large break LOCA (a leak before break)Duane Arnold estimated that this modification would cost at least $2M.$2,000,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-5441. Trip/Shutdown RiskReducing all initiating events except pipe breaks, floods, and LOSP by a factor of

28.0%6.7%6.9% $62,372 $187,11757 - Generation Risk Assessment implementation into plant activities (trip/

shutdown risk modeling).CNS estimate.$500,000Not cost effective42. Increase Availability of SSW Pump House Ventilation System Eliminated failure of SSW Pump House Ventilation1.6%2.2%2.3% $15,071 $45,21258 - Increase the training emphasis and provide additional control room indication on the operational status of SSW pump house ventilation system.

Hardware modification range

estimate.$100,000Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-5543. Increase recovery time of ECCS upon loss of

SSW Eliminated failure of SSW to the LPCS room cooler4.1%6.5%6.8% $40,452 $121,35759 - Increase operator training for alternating operation of the low pressure ECCS pumps (LPCI and LPCS) for loss of SSW scenarios.Procedure with training range

estimate.$50,000Retain44. Additional Containment Heat

Removal Eliminated failure of suppression pool cooling and containment spray systems27.5%53.2%56.3% $298,121 $894,36260 - Install an additional method of heat removal from containment.Plant-specific cost estimate.$4,352,023Not cost effectiveTable E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-5645. Improve RHR Heat Exchanger Availability Eliminated failure of RHR HX Cooler inlet and outlet valves3.6%7.8%8.3% $41,340 $124,01962 - Add a bypass around the RHR HX inlet and outlet valvesPlant-specific cost estimate.$2,831,652Not cost effective46. Improve RCIC Lube Oil Cooling Eliminated the failure to cool RCIC lube oil4.7%1.9%1.6% $30,894 $92,68363 - Add a redundant RCIC lube oil cooling path.Hardware modification range

estimate.$100,000Not cost effective1.These analysis cases only impact external events and have been evaluated differently as shown in Section E.2.3

.Table E.2-2 (Continued)Summary of Phase II SAMA Candidates Considered in Cost-Benefit EvaluationAnalysis Case (bold)SAMA Number and TitleAssumptions CDF Reduction PDR Reduction OECR Reduction Internal and External BenefitInternal and External Benefit with UncertaintyGGNS Cost EstimateConclusion Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-57Table E.2-3Sensitivity Analysis ResultsAnalysis Case (bold)SAMA Number and TitleInternal and External Benefit, 20 yrs Remaining, 7% Discount RateSensitivity Case 1, Internal and External Benefit, 33 yrs Remaining, 7% Discount RateSensitivity Case 2, Internal and External Benefit, 20 yrs Remaining, 3% Discount Rate GGNS Cost Estimate 1. DC Power $115,656 $171,775 $144,4231 - Provide additional DC battery capacity$500,0002 - Replace lead-acid batteries with fuel cells$1,000,00011 - Portable generator for direct current (DC) power: This SAMA involves the use of a portable generator to supply DC power to the battery chargers during a station blackout.$714,00012 - Portable generator for direct current (DC) power: This SAMA involves the use of a portable generator to supply DC power to the individual panels during a station blackout.$714,00015 - Use DC generators to provide power to operate the switchyard power control breakers while a 480-V AC generator could supply the air compressors for breaker support.$1,428,0002. Improve Charger Reliability $13,598 $19,619 $17,2763 - Add battery charger to existing DC system$90,00013 - Proceduralize battery charger high-voltage shutdown circuit inhibit$50,0003. Add DC System Cross-Ties $73,056 $105,875 $92,577 4 - Provide DC bus cross-ties$300,000 Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-584. Increase Availability of On-Site AC Power $149,396 $221,380 $186,8145 - Provide an additional diesel generator$20,000,0008 - Install a gas turbine generator with tornado protection$2,000,0005. Improve AC Power $177,524 $262,069 $222,495 6 - Improve 4.16-kV bus cross-tie ability$656,00017 - Provide alternate feeds to essential loads directly from an alternate emergency bus$656,0006. Reduce Loss of Off-Site Power During Severe Weather $26,087 $38,786 $32,5547 - Install an additional, buried off-site power source. $2,485,0007. Provide Backup EDG Cooling $16,515 $24,490 $20,6429 - Use fire water system as backup source for diesel cooling$100,00010 - Add new backup source of diesel cooling$2,000,0008. Increase EDG Reliability $30,348 $44,328 $38,27914 - Provide a portable EDG fuel oil transfer pump$100,000Table E.2-3 (Continued)Sensitivity Analysis ResultsAnalysis Case (bold)SAMA Number and TitleInternal and External Benefit, 20 yrs Remaining, 7% Discount RateSensitivity Case 1, Internal and External Benefit, 33 yrs Remaining, 7% Discount RateSensitivity Case 2, Internal and External Benefit, 20 yrs Remaining, 3% Discount Rate GGNS Cost Estimate Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-599. Improve DG reliability $2,181 $3,249 $2,71816 - Provide a diverse swing diesel generator air start compressor$100,00010. Reduce Plant-Centered Loss of Off-Site Power $76,556 $113,849 $95,52218 - Protect transformers from failure$780,00011. Redundant Power to Torus Hard Pipe Vent (THPV) Valves $10,766 $15,502 $13,69419 - Provide redundant power to direct torus hard pipe vent valves to improve the reliability of the direct torus vent valves and enhance the containment heat removal capability.$714,00012. High Pressure Injection System $594,912 $901,576 $733,64520 - Install an independent active or passive high pressure injection system$8,800,00061 - Install a backup water supply and pumping capability that is independent of normal and emergency AC power$6,409,949Table E.2-3 (Continued)Sensitivity Analysis ResultsAnalysis Case (bold)SAMA Number and TitleInternal and External Benefit, 20 yrs Remaining, 7% Discount RateSensitivity Case 1, Internal and External Benefit, 33 yrs Remaining, 7% Discount RateSensitivity Case 2, Internal and External Benefit, 20 yrs Remaining, 3% Discount Rate GGNS Cost Estimate Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-6013. Extend RCIC Operation $10,031 $14,448 $12,75721 - Raise HPCI/RCIC backpressure trip set points [HPCI backpressure trip setpoint has already been raised. This SAMA will evaluate raising the RCIC backpressure trip set point]. $200,00014. Improve ADS System $299,106 $469,925 $360,32022 - Modify automatic depressurization system components to improve reliability [This SAMA will add larger accumulators thus increasing reliability during SBOs].$1,176,85015. Improve ADS Signals $129,383 $205,503 $154,71923 - Add signals to open safety relief valves automatically in an MSIV closure transient. $1,500,00016. Low Pressure Injection System $229,965 $331,005 $292,57424 - Add a diverse low pressure injection system.$8,800,00017. ECCS Low Pressure Interlock $10,031 $14,448 $12,75725 - Install a bypass switch to allow operators to bypass the low reactor pressure interlock circuitry that inhibits opening the LPCI or core spray injection valves following sensor or logic failures that prevent all low pressure injection valves from opening.$1,000,000Table E.2-3 (Continued)Sensitivity Analysis ResultsAnalysis Case (bold)SAMA Number and TitleInternal and External Benefit, 20 yrs Remaining, 7% Discount RateSensitivity Case 1, Internal and External Benefit, 33 yrs Remaining, 7% Discount RateSensitivity Case 2, Internal and External Benefit, 20 yrs Remaining, 3% Discount Rate GGNS Cost Estimate Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-6118. RHR Heat Exchangers $205,223 $290,595 $263,55726 - Implement modifications to allow manual alignment of the fire water system to RHR heat exchangers.$1,950,00019. Emergency Service Water System Reliability $37,903 $54,031 $48,491 27 - Add a service water pump to increase availability of cooling water$5,900,00020. Main Feedwater System Reliability $162,050 $241,055 $202,16328 - Add a motor-driven feed water pump$1,650,00021. Increase Availability of Room Cooling $175,400 $265,739 $216,34229 - Provide a redundant train or means of ventilation$2,202,72522. Increase Availability of the DG System through HVAC Improvements $75,988 $113,283 $94,66930 - Add a diesel building high temperature alarm or redundant louver and thermostat.$1,304,70032 - Diverse EDG HVAC logic$300,00033 - Install additional fan and louver pair for EDG heating, ventilation, and air conditioning$6,000,000Table E.2-3 (Continued)Sensitivity Analysis ResultsAnalysis Case (bold)SAMA Number and TitleInternal and External Benefit, 20 yrs Remaining, 7% Discount RateSensitivity Case 1, Internal and External Benefit, 33 yrs Remaining, 7% Discount RateSensitivity Case 2, Internal and External Benefit, 20 yrs Remaining, 3% Discount Rate GGNS Cost Estimate Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-6223. Increased Reliability of HPCI and RCIC Room Cooling $10,031 $14,448 $12,75731 - Create ability to switch HPCI and RCIC room fan power supply to DC in an SBO event.$300,00024. Increase Reliability of Instrument Air $137,842 $201,172 $173,95134 - Modify procedure/hardware to provide ability to align diesel power to more air compressors$1,200,00035 - Replace service and instrument air compressors with more reliable compressors which have self-contained air cooling by shaft-driven fans$1,394,59826. Backup Nitrogen to SRV $40,614 $62,050 $49,82836 - Install nitrogen bottles as backup gas supply for safety relief valves.$1,722,70626. Improve Availability of SRVs and MSIVs $300,631 $472,257 $362,19037 - Improve SRV and MSIV pneumatic components.$1,500,00027. Improve Suppression Pool Cooling $205,223 $290,595 $263,55738 - Install an independent method of suppression pool cooling.$5,800,000Table E.2-3 (Continued)Sensitivity Analysis ResultsAnalysis Case (bold)SAMA Number and TitleInternal and External Benefit, 20 yrs Remaining, 7% Discount RateSensitivity Case 1, Internal and External Benefit, 33 yrs Remaining, 7% Discount RateSensitivity Case 2, Internal and External Benefit, 20 yrs Remaining, 3% Discount Rate GGNS Cost Estimate Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-6328. Increase Availability of Containment Heat Removal $288,437 $409,794 $369,72239 - Procedural change to cross-tie open cycle cooling system to enhance containment spray system$25,00041 - Use the fire water system as a backup source for the drywell spray system$1,950,00029. Decay Heat Removal Capability - Drywell Spray $288,550 $409,953 $369,86640 - Install a passive drywell spray system to provide redundant drywell spray method.$5,800,00030. Increase Availability of the CST$107,899$156,536$136,64342 - Enhance procedures to refill CST from demineralized water or service water system.$200,00031. Filtered Vent to Increase Heat Removal Capacity for Non-ATWS Events $80,920 $96,745 $113,07443 - Install a filtered containment vent to provide fission product scrubbing$1,500,00032. Reduce Hydrogen Ignition $142,455 $209,206 $179,10444 - Provide post-accident containment inerting capability.$2,665,12345 - Install a passive hydrogen control system.$760,000Table E.2-3 (Continued)Sensitivity Analysis ResultsAnalysis Case (bold)SAMA Number and TitleInternal and External Benefit, 20 yrs Remaining, 7% Discount RateSensitivity Case 1, Internal and External Benefit, 33 yrs Remaining, 7% Discount RateSensitivity Case 2, Internal and External Benefit, 20 yrs Remaining, 3% Discount Rate GGNS Cost Estimate Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-6433. Controlled Containment Venting $31,080 $44,274 $39,77946 - Provide passive overpressure relief by changing the containment vent valves to fail open and improving the strength of the rupture disk$1,000,00047 - Enable manual operation of all containment vent valves via local controls$150,000 34. ISLOCA $77 $118 $9548 - Increase frequency of valve leak testing to reduce ISLOCA frequency$100,00050 - Revise EOPs to improve ISLOCA identification$50,00051 - Improve operator training on ISLOCA coping$112,00035. MSIV Design $10,031 $14,448 $12,75749 - Improve MSIV design to decrease the likelihood of containment bypass scenarios.$1,000,00036. SLC System $10,616 $15,376 $13,45852 - Increase boron concentration in the SLC system [Reduced time required to achieve shutdown provides increased margin in the accident timeline for successful initiation of SLC]$50,000Table E.2-3 (Continued)Sensitivity Analysis ResultsAnalysis Case (bold)SAMA Number and TitleInternal and External Benefit, 20 yrs Remaining, 7% Discount RateSensitivity Case 1, Internal and External Benefit, 33 yrs Remaining, 7% Discount RateSensitivity Case 2, Internal and External Benefit, 20 yrs Remaining, 3% Discount Rate GGNS Cost Estimate Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-6537. SRV Reseat $29,108 $42,416 $36,76753 - Increase safety relief valve (SRV) reseat reliability to address the risk associated with dilution of boron caused by the failure of the SRVs to reseat after standby liquid control (SLC) injection$2,200,00038. Add Fire Suppression 1N/AN/AN/A54 - Add automatic fire suppression systems to the dominant fire zones$375,00039. Reduce Risk from Fires that Require Control Room Evacuation 1N/AN/AN/A55 - Upgrade the ASDS panel to include additional system controls for opposite division.$786,99140. Large Break LOCA $316,124 $463,652 $380,82756 - Provide digital large break LOCA protection to identify symptoms/precursors of a large break LOCA (a leak before break)$2,000,00041. Trip/Shutdown Risk $62,372 $94,032 $77,17057 - Generation Risk Assessment implementation into plant activities (trip/shutdown risk modeling).$500,000Table E.2-3 (Continued)Sensitivity Analysis ResultsAnalysis Case (bold)SAMA Number and TitleInternal and External Benefit, 20 yrs Remaining, 7% Discount RateSensitivity Case 1, Internal and External Benefit, 33 yrs Remaining, 7% Discount RateSensitivity Case 2, Internal and External Benefit, 20 yrs Remaining, 3% Discount Rate GGNS Cost Estimate Grand Gulf Nuclear StationApplicant's Environmental ReportOperating License Renewal Stage E.2-6642. Increase Availability of SSW Pump House Ventilation System $15,071 $21,998 $19,01758 - Increase the training emphasis and provide additional control room indication on the operational status of SSW pump house ventilation system.$100,00043. Increase Recovery Time of ECCS upon Loss of SSW $40,452 $58,438 $51,35759 - Increase operator training for alternating operation of the low pressure ECCS pumps (LPCI and LPCS) for loss of SSW scenarios.$50,00044. Additional Containment Heat Removal $298,121 $423,739 $382,03860 - Install an additional method of heat removal from containment.$4,352,02345. Improve RHR Heat Exchanger Availability $41,340 $58,200 $53,26362 - Add a bypass around the RHR HX inlet and outlet valves$2,831,65246. Improve RCIC Lube Oil Cooling $30,894 $48,447 $37,26463 - Add a redundant RCIC lube oil cooling path.$100,0001.These analysis cases only impact external events and have been evaluated differently as shown in Section E.2.3

.Table E.2-3 (Continued)Sensitivity Analysis ResultsAnalysis Case (bold)SAMA Number and TitleInternal and External Benefit, 20 yrs Remaining, 7% Discount RateSensitivity Case 1, Internal and External Benefit, 33 yrs Remaining, 7% Discount RateSensitivity Case 2, Internal and External Benefit, 20 yrs Remaining, 3% Discount Rate GGNS Cost Estimate

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