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

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Attachment E Severe Accident Mitigation Alternatives Analysis Attachment E contains the following sections.

E.1 - Evaluation of GGNS PSA Model E.2 - Evaluation of GGNS SAMA Candidates E-1

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage TABLE OF CONTENTS E.1 EVALUATION OF PROBABILISTIC SAFETY ANALYSIS MODEL . . . . . . . . . . . . . .E.1-1 E.1.1 PSA Model - Level 1 Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-1 E.1.2 PSA Model - Level 2 Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-23 E.1.2.1 Containment Performance Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-23 E.1.2.2 Radionuclide Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-48 E.1.2.2.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-48 E.1.2.2.2 Timing of Release. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-48 E.1.2.2.3 Magnitude of Release. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-48 E.1.2.2.4 Release Category Bin Assignments . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-49 E.1.2.2.5 Mapping of Level 1 Results into the Various Release Categories . . . . . .E.1-50 E.1.2.2.6 Process Used to Group the Source Terms . . . . . . . . . . . . . . . . . . . . . . . .E.1-54 E.1.2.2.7 Consequence Analysis Source Terms . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-56 E.1.2.2.8 Release Magnitude Calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-56 E.1.3 IPEEE Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-59 E.1.3.1 Seismic Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-59 E.1.3.2 Fire Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-59 E.1.3.3 Other External Hazards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-67 E.1.4 PSA Model Revisions and Peer Review Summary . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-68 E.1.4.1 Major Differences between the 1997(R1) PSA Model and the IPE Model . . . . .E.1-68 E.1.4.2 Major Differences between the 2002 (R2) PSA Model and the 1997(R1)

PSA Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-68 E.1.4.3 Major Differences between the 2010 (R3) PSA Model and the 2002 (R2)

PSA Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-69 E.1.4.4 Major Differences between the 2010 (EPU) PSA Model and the 2010 (R3)

PSA Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-69 E.1.4.5 PSA Model Peer Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-71 E.1.5 The MACCS2 ModelLevel 3 Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-71 E.1.5.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-71 E.1.5.2 Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-72 E.1.5.2.1 Projected Total Population by Spatial Element. . . . . . . . . . . . . . . . . . . . .E.1-72 E.1.5.2.2 Land Fraction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-74 E.1.5.2.3 Watershed Class . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-74 E.1.5.2.4 Regional Economic Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-74 E.1.5.2.5 Agriculture Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-75 E.1.5.2.6 Meteorological Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-75 E.1.5.2.7 Emergency Response Assumptions. . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-76 E.1.5.2.8 Core Inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-76 E.1.5.2.9 Source Terms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-78 E.1.5.3 RESULTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-78 E.1.6 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.1-81 i

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.2 EVALUATION OF GGNS SAMA CANDIDATES. . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.2-1 E.2.1 SAMA List Compilation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.2-1 E.2.2 Qualitative Screening of SAMA Candidates (Phase I) . . . . . . . . . . . . . . . . . . . . . . . .E.2-2 E.2.3 Final Screening and Cost Benefit Evaluation of SAMA Candidates (Phase II) . . . . .E.2-2 E.2.4 Sensitivity Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.2-15 E.2.5 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .E.2-17 ii

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage LIST OF TABLES Table E.1-1 GGNS EPU Model CDF Results by Major Initiators . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-3 Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF) . . E.1-5 Table E.1-3 Notation and Definitions for GGNS CET Functional Nodes Description . . . . . . . . . . . . . E.1-25 Table E.1-4 Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-40 Table E.1-6 GGNS Release Categories . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-50 Table E.1-5 Release Severity and Timing Classification Scheme Summary . . . . . . . . . . . . . . . . . . . E.1-50 Table E.1-7 Summary of GGNS Core Damage Accident Sequences Plant Damage States . . . . . . . E.1-52 Table E.1-8 Summary of Containment Event Tree Quantification . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-55 Table E.1-9 GGNS Release Category Source Terms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-57 Table E.1-10 GGNS Fire IPEEE Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-61 Table E.1-11 Estimated Population Distribution within a 50-Mile Radius . . . . . . . . . . . . . . . . . . . . . . . E.1-73 Table E.1-12 Estimated GGNS Core Inventory (Becquerels). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-77 Table E.1-13 Base Case Mean PDR and OECR Values for Postulated Internal Events . . . . . . . . . . . E.1-79 Table E.1-14 Summary of Offsite Consequence Results for Sensitivity Results . . . . . . . . . . . . . . . . . E.1-80 iii

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-1 Phase I SAMAs Related to IPE and IPEEE Insights. . . . . . . . . . . . . . . . . . . . . . . . . . . . E.2-19 Table E.2-2 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation . . . . . E.2-30 Table E.2-3 Sensitivity Analysis Results. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.2-57 iv

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage LIST OF FIGURES Figure E.1-1 GGNS Radionuclide Release Category Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.1-38 Figure E.1-2 Summary of GGNS Core Damage Accident Sequences Plant Damage States . . . . . . . E.1-39 v

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage LIST OF ACRONYMS IN ATTACHMENTS E.1 AND E.2 Acronym Definition AC Alternating current ADS Automatic depressurization system ASDS Alternate shutdown system ATWS Anticipated transient without scram BWR Boiling water reactor BWROG Boiling Water Reactor Owners Group CCF Common cause failure CCW Component cooling water CDF Core damage frequency CET Containment event tree CNS Cooper Nuclear Station CPI Consumer Price Index CRD Control rod drive CsI Cesium iodide CST Condensate storage tank DC Direct current DF Decontamination factor DG Diesel generator ECCS Emergency core cooling system EDG Emergency diesel generator EOP Emergency operating procedure EPG Emergency procedure guidelines EPU Extended power uprate EPZ Emergency planning zone vi

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Acronym Definition ESF Emergency safety feature FIVE Fire induced vulnerability evaluation FPS Fire protection system FPW Fire protection water FW Feedwater GGNS Grand Gulf Nuclear Station HPCI High pressure coolant injection HPCS High pressure core spray HVAC Heating, ventilation and air conditioning IA Instrument air IPE Individual Plant Examination IPEEE Individual Plant Examination of External Events ISLOCA Interfacing systems loss of coolant accident LERF Large early release frequency LOCA Loss of coolant accident LOSP Loss of off-site power LPCI Low pressure coolant injection LPCS Low pressure core spray MAAP Modular Accident Analysis Program MACCS2 MELCOR Accident Consequences Code System 2 MSIV Main steam isolation valve MST Main steam tunnel NRC Nuclear Regulatory Commission OECR Off-site economic cost risk OSP Off-site power PCPL Primary containment pressure limit vii

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Acronym Definition PCS Power conversion system PDR Population dose risk PDS Plant damage state PHV Pump house ventilation PRA Probabilistic Risk Assessment PSA Probabilistic Safety Assessment PSW Plant service water RB Reactor building RCIC Reactor core isolation cooling RCS Reactor coolant system RHR Residual heat removal RPV Reactor pressure vessel RRW Risk reduction worth SAG Severe accident guideline SAMA Severe accident mitigation alternative SBO Station blackout SLC Standby liquid control SORV Stuck open relief valve SPC Suppression pool cooling SPMU Suppression pool makeup SRV Safety relief valve SSW Standby service water TBCW Turbine building cooling water WW Wetwell viii

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage ATTACHMENT E.1 EVALUATION OF GGNS PSA MODEL

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.1 EVALUATION OF PROBABILISTIC SAFETY ANALYSIS MODEL The 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.1.1 PSA Model - Level 1 Analysis The 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 Table E.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 0.5 percent and are judged to be highly unlikely contributors for the identification of cost-beneficial enhancements. These basic eventsincluding component failures, operator actions, and initiating eventswere reviewed to determine if additional SAMA actions may need to be considered.

E.1-1

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.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.

E.1-2

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-1 GGNS EPU Model CDF Results by Major Initiators Initiating Event Group Total IE Group Probability % CDF Large Loss of Coolant Accident 1.45E-07 7.10 (LOCA)

Feedwater Line Break Outside of 2.76E-10 0.00 Containment Plant Service Water (PSW) 1.00E-09 0.00 Flooding Initiator Reactor Vessel Rupture 1.00E-08 0.50 Intermediate LOCA 2.03E-08 1.00 Small LOCA 1.33E-11 0.00 Small-Small LOCA 2.47E-11 0.00 Standby Service Water (SSW) 6.55E-12 0.00 Flooding Initiator Loss of Off-Site Power Initiator 2.87E-07 14.00 Loss of 500 kV Power 5.12E-11 0.00 (Preferred)(1)

Loss of Power Conversion System 2.31E-07 11.20 (PCS) Initiator Closure of Main Steam Isolation 8.81E-08 4.30 Valves (MSIVs) (Initiator)

PCS Available Transient 6.32E-07 30.80 Loss of Condensate Feed Water 2.20E-07 10.70 Pumps Inadvertent Open Relief Valve 9.78E-09 0.50 Loss of Alternating Current (AC) 1.79E-08 0.90 Division 1 Initiator Loss of AC Division 2 Initiator 3.82E-08 1.90 Loss of Turbine Cooling Water 8.09E-09 0.40 (TBCW)

Loss of Component Cooling Water 6.87E-10 0.00 Initiating Event Loss of Control Rod Drive (CRD) 2.20E-09 0.10 E.1-3

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-1 (Continued)

GGNS EPU Model CDF Results by Major Initiators Initiating Event Group Total IE Group Probability % CDF Loss of Direct Current (DC) 2.22E-10 0.00 Division 1 Initiator Loss of DC Division 2 Initiator 1.30E-10 0.00 Loss of Instrument Air 1.36E-07 6.60 Loss of PSW Initiating Event 1.50E-09 0.10 Loss of Service Transformer 11 9.20E-08 4.50 Loss of Service Transformer 21 1.09E-07 5.30 Interfacing System Loss of 2.03E-10 0.00 Coolant Accident (ISLOCA) in Shutdown Cooling Supply Header (Pen 14)

Total CDF 2.05E-06 100.00 Total Anticipated Transient without ~ 3.08E-09 0.15 Scram (ATWS)(2)

Total Station Blackout (SBO)(2) ~ 7.51E-07 36.65 (TB)

1. 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.

E.1-4

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition

%A 3.19E-05 1.036 Large LOCA This term represents a large LOCA. Phase II SAMA 56 for detecting LOCAs was evaluated.

%S1 4.69E-06 1.005 Intermediate LOCA This term represents an intermediate LOCA. Phase II SAMA 57 for implementing a GRA was evaluated.

%T1 2.48E-02 1.6289 Loss of offsite power initiator This term represents a loss of offsite power initiator. Phase II SAMAs 7, 15 and 18 for improving offsite, switchyard and transformer availability were evaluated.

%T2 1.77E-01 1.0969 Loss of PCS initiator This term represents a loss of power conversion system initiator. Phase II SAMA 28 for improving availability and reliability of feedwater was evaluated.

%T2M 2.01E-01 1.0423 Closure 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.

%T3A 7.98E-01 1.2633 PCS available transient This 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.

%T3B 2.00E-01 1.0934 Loss of condensate feed This term represents a loss of condensate feedwater pumps water pumps initiator. Phase II SAMA 28 for improving availability and reliability of feedwater was evaluated.

E.1-5

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition

%TAC1 2.56E-03 1.0078 Loss of AC Division 1 This term represents the loss of AC Division 1 initiator. Phase II initiator SAMAs 5 and 8 for enhancing AC system reliability or to cope with loss of offsite power and SBO events were evaluated.

%TAC2 2.56E-03 1.0191 Loss of AC Division 2 This term represents the loss of AC Division 1 initiator. Phase II initiator SAMAs 5 and 8 for enhancing AC system reliability or to cope with loss of offsite power and SBO events were evaluated.

%TIA 3.51E-03 1.0487 Loss of instrument air This term represents a loss of instrument air initiator. Phase II SAMAs 34, 35, 36, and 37 for improving the instrument air system were evaluated.

%TST11 9.85E-02 1.029 Loss of service transformer This term represents a loss of service transformer 11. Phase II 11 SAMA 18 for protecting transformers was evaluated.

%TST21 7.48E-02 1.0384 Loss of service transformer This term represents a loss of service transformer 21. Phase II 21 SAMA 18 for protecting transformers was evaluated.

B21-FO- 1.00E+00 1.0629 Operator fails to connect gas This term represents a failure to manually operate ADS when IA HEBOTTLES bottles to ADS air header is lost. Phase II SAMA 36 for adding automatic nitrogen backup to ADS components was evaluated.

B21-FO-HEDEP2-I 1.00E+00 1.5587 Operator fails to manually This term represents a failure to manually operate ADS when IA depressurize vessel with is lost. Phase II SAMAs 20, 21, 22, and 28 for improving high non-ADS valves pressure injection or ADS components were evaluated.

E12-CF-MVLPCS 5.73E-05 1.0072 Two or more LPSI and LPCS This term represents a failure of LPCI injection valves to open.

injection MOVs to open Phase II SAMA 25 for bypassing LPCI low pressure permissives was evaluated.

E12-LF-FGCS 1.00E+00 1.0637 Containment spray signal This term is a flag. No SAMAs need to be aligned.

generated E.1-6

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition E12-MA-TMRHRA 8.83E-03 1.0118 RHR Train A unavailable This term represents a failure of LPCI. Phase II SAMAs 24, and due to maintenance 25 for improving or adding low pressure injection systems were evaluated.

E12-MA-TMRHRB 5.56E-03 1.0066 RHR Train B unavailable This term represents a failure of LPCI. Phase II SAMAs 24, and due to maintenance 25 for improving or adding low pressure injection systems were evaluated.

E22-042-H 6.40E-03 1.0377 Suppression pool suction This term represents a failure of a high pressure injection.

line hardware failure (long Phase II SAMAs 20, 21, 22, and 28 for improving or adding high term) pressure injection were evaluated.

E22-CC-MVF004-G 6.30E-03 1.0742 Normally closed motor This term represents a failure of a high pressure injection.

driven valve FOO4 fails to Phase II SAMAs 20, 21, 22, and 28 for improving or adding high open pressure injection were evaluated.

E22-CC-MVF012-G 6.30E-03 1.0742 Minimum flow valve F012-C This term represents a failure of a high pressure injection.

fails to open Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E22-FS-MPC001-G 3.00E-03 1.0336 HPCS motor driven pump This term represents a failure of a high pressure injection.

C001 fails to start Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E22-HW-ICHPCS-G 1.60E-03 1.0174 HPCS actuation circuit This term represents a failure of a high pressure injection.

failure Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E22-HW-ICMNFLO- 1.60E-03 1.0174 Minimum flow control circuit This term represents a failure of a high pressure injection.

G failure Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E.1-7

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition E22-MA-MAHPCS- 6.59E-03 1.0662 HPCS unavailable due to This term represents a failure of a high pressure injection.

G maintenance Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E22-OO-MVF012-G 3.40E-03 1.0384 Normally open motor driven This term represents a failure of a high pressure injection.

valve E22-F012 fails to close Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E51-021-H 7.99E-03 1.0127 Suppression pool suction This term represents a failure of a high pressure injection.

switchover fails due to Phase II SAMAs 20, 21, 22, and 28 for improving or adding high hardware (long term) pressure injection were evaluated.

E51-026-G 6.40E-03 1.0634 RCIC pump failsminimum This term represents a failure of a high pressure injection.

flow path fails to open Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E51-035M-G 1.27E-02 1.1348 RCIC steam supply valves This term represents a failure of a high pressure injection.

fail Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E51-043-G 8.29E-03 1.0839 Lube oil cooling line This term represents a failure of a high pressure injection.

hardware failure Phase II SAMA 63 for improving RCIC reliability was evaluated.

E51-CC-MVF013A- 6.30E-03 1.0623 Motor-operated valve F013- This term represents a failure of a high pressure injection.

G A fails to open Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E51-FF-FSC001-G 3.51E-03 1.0335 RCIC pump start failures 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.

E.1-8

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition E51-FO-HEF031A- 1.00E+00 1.0063 Operator fails to open SP This term represents a failure of a high pressure injection.

G suction valve F031-A Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E51-FO- 1.00E+00 1.0307 Human error: Failure to This term represents a failure of a high pressure injection.

HETRPBYP bypass RCIC temperature Phase II SAMAs 20, 21, 22, and 28 for improving or adding high trips (EOP Attachment 3) pressure injection were evaluated.

E51-FR-TPC001-G 2.01E-01 1.2014 RCIC turbine-driven pump This term represents a failure of a high pressure injection.

fails to run Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E51-FR-TPC18HR- 6.71E-02 1.0586 RCIC turbine fails to run for 8 This term represents a failure of a high pressure injection.

G hours Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E51-HW-ICLVL8-I 1.60E-03 1.0148 Hardware failure of level 8 This term represents a failure of a high pressure injection.

isolation circuitry Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E51-HW-ICSYACT- 1.60E-03 1.0148 RCIC actuation circuitry This term represents a failure of a high pressure injection.

G failure Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

E51-MA-TMRCIC-G 1.24E-02 1.1075 RCIC unavailable due to This term represents a failure of a high pressure injection.

maintenance Phase II SAMAs 20, 21, 22, and 28 for improving or adding high pressure injection were evaluated.

HVC-LF- 1.00E+00 1.0113 Failure of SSW A pump This term is a flag. No SAMAs need to be aligned.

FGSSWAPH house ventilation E.1-9

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition L21-OP-BT-1A3-D 1.00E+00 1.0382 Battery 1A3 discharged This term represents battery depletion before recovery of offsite

(~ 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months depletion time) power. Phase II SAMAs 1, 2, 3, 11,12, and 15 for extending available recovery time by improving DC power were evaluated.

LOSP-EPRI 1.00E-03 1.022 Conditional LOSP after a This term represents a transient induced loss of offsite power.

plant trip Phase II SAMAs 5 and 8 for improving AC power reliability were evaluated.

M24-RP- 9.38E-03 1.0714 ECCS pump failure due to This term represents loss of ECCS equipment due to CTFLECCS containment failure 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- 7.98E-03 1.0099 Hardware failure of the This term represents a failure of the containment vent valves.

MLVNTHW-Q containment venting 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-I 1.00E+00 1.0827 Human error: Failure to This term represents a failure of a human action to restore restart reactor feed pumps feedwater and manually depressurize. Phase II SAMAs 20 and following level 9 trip 61 for improving high pressure injection capability were evaluated.

N21-FO-HEPCS-G 1.00E+00 1.1081 Human error: Failure to This term represents a failure of a human action to restore properly align the PCS for feedwater and manually depressurize. Phase II SAMAs 20 and injection 40 for improving high pressure injection and suppression pool cooling capability were evaluated.

E.1-10

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition NR-ACHWR-1HRS 6.00E-01 1.011 Failure to recover AC bus This term represents a failure to recover the AC bus. Phase II failure in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SAMAs 5, 8, 17 and 18 for protecting or providing alternate bus power supplies were evaluated.

NR-ACHWR-8HRS 1.00E-02 1.0158 Failure to recover AC bus This term represents a failure to recover the AC bus. Phase II failure in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months SAMAs 5, 8, 17 and 18 for protecting or providing alternate bus power supplies were evaluated.

NRC-DEP-RCIC 8.40E-03 1.0051 Failure to manually This term represents a failure of a human action to manually depressurize using RCIC depressurize using RCIC. Phase II SAMA 22 increased ADS reliability was evaluated.

NRC-DG-CF1HRS 9.00E-01 1.0059 Failure to recover diesel This term represents a failure of a human action to recover the generator common cause DG common cause failure in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Phase II SAMAs 5 and 8 failure in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to install an additional diesel or gas turbine generator were evaluated.

NRC-DGHW10&FW 2.85E-01 1.0085 Failure to recover DG This term represents a failure of a human action to recover DG hardware failure or start FW hardware failure or start FW in 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . Phase II SAMAs 5 and in 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months 8 to install an additional diesel or gas turbine generator were evaluated.

NRC-DG- 5.00E-01 1.005 Failure to recover diesel This term represents a failure of a human action to recover DG HW10HRS generator hardware failure in hardware failure in 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . Phase II SAMAs 5 and 8 to install 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months an additional diesel or gas turbine generator were evaluated.

NRC-DG-HW1HR 9.00E-01 1.0107 Failure to recover diesel This term represents a failure of a human action to recover DG generator hardware failure in hardware failure in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Phase II SAMAs 5 and 8 to install an 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months additional diesel or gas turbine generator were evaluated.

E.1-11

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition NRC-DG-MA1HR 9.00E-01 1.0197 Failure to recover diesel This term represents a failure of a human action to recover DG generator from maintenance hardware failure in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Phase II SAMAs 5 and 8 to install an in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months additional diesel or gas turbine generator were evaluated.

NRC-FO- 1.30E-03 1.0447 Failure to connect air bottles This term represents a failure of a human action to connect air ADSBOTTLE to SRV accumulators bottles to the SRV accumulators. Phase II SAMAs 22 and 37 to add larger accumulators and improve SRV pneumatic components were evaluated.

NRC-FO-FWS8HR 1.10E-02 1.0107 Failure to align FPW for long This term represents a failure of a human action to align the term injection firewater system for injection. Phase II SAMAs 24 and 25 for Improved low pressure injection capability were evaluated.

NRC-FO-FWSACT 5.70E-01 1.0927 Failure to align FPW for long This term represents a failure of a human action to align the term injection firewater system for injection. Phase II SAMAs 24 and 25 for Improved low pressure injection capability were evaluated.

NRC-OSP-CNT 1.21E-02 1.0052 Fail to recover OSP given This term represents a failure to recover offsite power. Phase II long term containment failure SAMAs 1, 2, 3, 11, 12, 14 and 15 for extending available recovery time by improving DC power were evaluated.

NRC-OSP-DLG0 1.28E-01 1.0135 Fail to recover OSP given This term represents a failure to recover offsite power. Phase II 0 FTR

No SSW PHV SAMAs 1, 2, 3, 11, 12, 14 and 15 for extending available failures recovery time by improving DC power were evaluated.

NRC-OSP-DSG0 6.18E-01 1.3513 Fail to recover OSP given U2 This term represents a failure to recover offsite power. Phase II

0 FTR

No SSW PHV SAMAs 1, 2, 3, 11, 12, 14 and 15 for extending available failures recovery time by improving DC power were evaluated.

NRC-OSP- 2.62E-01 1.0058 Fail to recover OSP given U2 This term represents a failure to recover offsite power. Phase II DSG0SSW0

0 FTR

1 or 2 SSW PHV SAMAs 1, 2, 3, 11, 12, 14 and 15 for extending available FTS recovery time by improving DC power were evaluated.

E.1-12

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition NRC-OSP-DSG1 1.05E-01 1.0855 Fail to recover OSP given U2 This term represents a failure to recover offsite power. Phase II

1 FTR

No SSW PHV SAMAs 1, 2, 3, 11, 12, 14 and 15 for extending available failures recovery time by improving DC power were evaluated.

NRC-OSP-DSG2 4.53E-02 1.0126 Fail to recover OSP given U2 This term represents a failure to recover offsite power. Phase II

2 FTR

No SSW PHV SAMAs 1, 2, 3, 11, 12, 14 and 15 for extending available failures recovery time by improving DC power were evaluated.

NRC-OSP-PSG0 7.63E-01 1.0134 Fail to recover OSP given This term represents a failure to recover offsite power. Phase II SRV LOCA

U2

0 FTR

SAMAs 1, 2, 3, 11, 12, 14 and 15 for extending available No SSW PHV failures recovery time by improving DC power were evaluated.

NRC-SSWPH-VENT 3.80E-04 1.0054 Failure to install alternate This term represents a failure of a human action to install means of cooling to SSW alternate means of cooling to the SSW pump house. Phase II pump house SAMA 58 for increasing training emphasis and providing control room indication on status of the SSW pump house HVAC was evaluated.

NR-PCS-60MN 6.00E-01 1.0402 Failure to recover PCS in This term represents a failure of a human action to restore 60 minutes feedwater and manually depressurize. Phase II SAMA 20 for improving high pressure injection capability was evaluated.

NRS-ALT-PWR- 4.50E-04 1.0068 Failure to align alternate This term represents a failure of a human action to align SUP power to 4.16 kV or 6.9 kV alternate power to 4.16 kV or 6.9 kV buses. Phase II SAMA 6 buses for improving 4.16kV bus cross-tie ability was evaluated.

NRS-DEP-LONG 1.20E-05 1.1703 Failure to manually This term represents a failure of a human action to manually depressurize with ADS/ depressurize with ADS/SRVs after more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Phase II SRVs (after more than 2 SAMAs 22 and 37 to add larger accumulators and improve SRV hours) pneumatic components were evaluated.

E.1-13

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition NRS-DEP-SHORT 3.20E-04 1.1895 Failure to manually This term represents a failure of a human action to manually depressurize with ADS/ depressurize with ADS/SRVs. Phase II SAMAs 22 and 37 to SRVs add larger accumulators and improve SRV pneumatic components were evaluated.

NRS-FO-SSWIA 2.20E-04 1.0104 Failure to align SSW B to This term represents a failure of a human action to align SSW B cool IA compressors to cool IA compressors. Phase II SAMA 35 for adding IA compressors which do not require cooling was evaluated.

NRS-PCS&DEP 4.20E-05 1.037 Failure to restore feedwater This term represents a failure of a human action to restore and manually depressurize feedwater and manually depressurize. Phase II SAMA 20 for improving high pressure injection capability was evaluated.

NRS-PCSL8&DEP 1.70E-05 1.0146 Failure to restore feedwater This term represents a failure of a human action to restore and manually depressurize feedwater and manually depressurize. Phase II SAMA 20 for improving high pressure injection capability was evaluated.

NRS-Y47&FPW 2.20E-04 1.0074 Failure of SSW ventilation This term represents a failure of a human action to restore SSW and align FPW ventilation and align FPW. Phase II SAMA 58 for increased training on restoring SSW ventilation and aligning FPW was evaluated.

OSP-LF-EVENTU2 1.00E+00 7.1015 RCIC failure This term is a flag. No SAMAs need to be aligned P1 1.13E-02 1.0087 One stuck-open relief valve This term represents stuck-open safety relief valves. Phase II SAMA 53 for increased SRV seating reliability was evaluated.

P11-PG-XVF021-G 7.20E-05 1.007 CST suction manual valve This term represents a blocked suction for both HPCS and P11-F021 plugs RCIC. Phase II SAMA 20 for adding alternate high pressure injection systems was evaluated.

E.1-14

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition P2 1.52E-03 1.0285 Two or more stuck-open This term represents stuck-open safety relief valves. Phase II relief valves SAMA 53 for increased SRV seating reliability was evaluated.

P41-004-A 6.43E-03 1.0096 Hardware failure of DG A This term represents a failure of cooling water to EDGs. Phase jacket cooler components II SAMAs 21 and 22 for adding a backup source of diesel cooling were evaluated.

P41-054-B 6.43E-03 1.0092 Hardware failure of DG B This term represents a failure of cooling water to EDGs. Phase jacket cooler components II SAMAs 9 and 10 for adding a backup source of diesel cooling were evaluated.

P41-152-L 6.87E-03 1.0094 Hardware failure of RHR This term represents a failure of the train A RHR heat heat exchanger coolers exchanger coolers or isolation valves. Phase II SAMA 62 for Train A bypassing the RHR HX SSW isolation valves was evaluated.

P41-CC-MVF001A- 6.30E-03 1.0273 Normally closed motor This term represents a failure of cooling water to ECCS and R operated valve F001A fails PCS. Phase II SAMAs 26 and 27 for improving service water to to open ECCS and PCS were evaluated.

P41-CC-MVF001B- 6.30E-03 1.0217 Normally closed motor This term represents a failure of cooling water to ECCS and R driven valve F001B fails to PCS. Phase II SAMAs 26 and 27 for improving service water to open ECCS and PCS were evaluated.

P41-CC-MVF005A- 6.30E-03 1.0273 Normally closed motor This term represents a failure of cooling water to ECCS and R driven valve F005A fails to PCS. Phase II SAMAs 26 and 27 for improving service water to open ECCS and PCS were evaluated.

P41-CC-MVF005B- 6.30E-03 1.0218 Normally closed motor This term represents a failure of cooling water to ECCS and R driven valve F005B fails to PCS. Phase II SAMAs 26 and 27 for improving service water to open ECCS and PCS were evaluated.

E.1-15

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition P41-CC-MVF014B- 6.30E-03 1.0091 Motor operated valve This term represents a failure of the RHR HX SSW isolation L F014B-B fails to open valves. Phase II SAMA 62 for bypassing the RHR HX SSW isolation valves was evaluated.

P41-CF-CVDISCH- 1.02E-05 1.0057 Common cause failure of This term represents a failure of cooling water to ECCS and R SSW discharge check valves PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-CF-FNC003S- 3.66E-05 1.013 CCF of 3 or more SSW This term represents a failure of cooling water to ECCS and R cooling tower fans to start PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-CF-MVDISCH- 6.99E-05 1.0407 CCF of SSW discharge This term represents a failure of cooling water to ECCS and R MOVs FOO5B, FOO5A, & PCS. Phase II SAMAs 26 and 27 for improving service water to F011C to open ECCS and PCS were evaluated.

P41-CF-MVF001AB 1.85E-04 1.0117 CCF of isolation valves This term represents a failure of cooling water to ECCS and F001A and B to open PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-CF-MVF005AB 1.85E-04 1.0069 CCF of discharge MOVs This term represents a failure of cooling water to ECCS and F005A and B to open PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-CF-MVF14AB- 1.85E-04 1.0056 CCF of 2 of 2 SSW RHR HX This term represents a failure of the RHR HX SSW isolation L valves to open valves. Phase II SAMA 62 for bypassing the RHR HX SSW isolation valves was evaluated.

P41-CF-MVF68AB- 1.85E-04 1.0056 CCF of 2 of 2 SSW RHR HX This term represents a failure of the RHR HX SSW isolation L valves to open valves. Phase II SAMA 62 for bypassing the RHR HX SSW isolation valves was evaluated.

E.1-16

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition P41-CF-ST-SUCT-R 1.40E-05 1.008 2 of 2 SSW suction strainers This term represents a failure of cooling water to ECCS and CCF to plug PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-FF-MLABST-R 1.98E-04 1.0124 Common cause start failures This term represents a failure of cooling water to ECCS and of SSW pumps A & B PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-FF-MLC002C- 9.03E-04 1.0099 Train C pump start failures This term represents a failure of cooling water to ECCS and R PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-FF-MLTCVLV- 6.34E-03 1.0777 SSW Train C common valve This term represents a failure of cooling water to ECCS and R hardware failures PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-FR-MPC002C- 7.20E-04 1.0063 Motor driven pump C002C This term represents a failure of cooling water to ECCS and R fails to continue running PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-LF- 1.00E+00 1.0207 Logic flagSSW cooling This term is a flag. No SAMAs need to be aligned.

FNSSWABC tower fans fail P41-MA-SSWA-R 2.53E-03 1.0069 SSW Train A unavailable This term represents a failure of cooling water to ECCS and due to maintenance PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-MA-SSWB-R 3.42E-03 1.0093 SSW Train B unavailable This term represents a failure of cooling water to ECCS and due to maintenance PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

E.1-17

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition P41-MA-SSWC-R 3.84E-03 1.0386 SSW Train C unavailable This term represents a failure of cooling water to ECCS and due to maintenance PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P41-PG-ST- 1.39E-04 1.0089 Suct. source failure of motor This term represents a failure of cooling water to ECCS and SUCTAR pumps A & C PCS. Phase II SAMAs 26 and 27 for improving service water to ECCS and PCS were evaluated.

P53-FO- 1.00E+00 1.011 Operator fails to align SSW- This term represents a failure of a human action to align SSW-B HECOOLIAS B to IAS compressor upon to the IAS compressor upon loss of TBCW. Phase II SAMA 35 loss of TBCW for adding IA compressors which do not require cooling was evaluated.

P64-FO-HE-G 1.00E+00 1.1242 Operator fails to align This term represents a failure of a human action to align the firewater system for injection firewater system for injection. Phase II SAMAs 24 and 25 for improved low pressure injection capability were evaluated.

P64-LF-FGSHORT 1.00E+00 1.0471 Flag for transient sequences This term is a flag. No SAMAs need to be aligned.

utilizing firewater P75-CF-3DGR-Z 2.16E-04 1.007 CCF of all 3 EDGs to run This term represents a common cause failure to run of 3 EDGs.

Phase II SAMAs 5, 8, 9, 10, and 14 for improving EDG reliability or adding additional onsite power sources were evaluated.

P75-CF-3DGS-Z 1.53E-05 1.0054 CCF of all 3 EDGs to start This 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- 4.69E-02 1.0124 DG11 fails to run This term represents a failure of DG11 to run. Phase II SAMAs A 5, 8, 9, 10, 14, and 15 for improving EDG reliability or adding additional onsite power sources were evaluated.

E.1-18

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition P75-FR-DG-DG12- 4.69E-02 1.016 DG12 fails to run This term represents a failure of DG11 to run. Phase II SAMAs B 5, 8, 9, 10, 14 and 15 for improving EDG reliability or adding additional onsite power sources were evaluated.

P75-FS-DG-DG11-A 6.94E-03 1.0094 DG11 fails to start This 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-B 6.94E-03 1.0092 DG12 fails to start This 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-A 1.34E-02 1.0099 DG11 in maintenance This term represents a failure of DG11 to run. Phase II SAMAs outage 5, 8, 9, 10, 14 and 15 for improving EDG reliability or adding additional onsite power sources were evaluated.

P75-MA-DGDG12-B 1.19E-02 1.013 DG12 in maintenance This term represents a failure of DG11 to run. Phase II SAMAs outage 5, 8, 9, 10, 14 and 15 for improving EDG reliability or adding additional onsite power sources were evaluated.

P81-FR-DG-DG13- 4.66E-02 1.0184 DG13 fails to run This term represents a failure DG13 to run. Phase II SAMAs 5, C 8, 9, 10, and 14 for improving EDG reliability or adding additional onsite power sources were evaluated.

P81-FS-DG-DG13- 5.97E-03 1.0101 DG13 fails to start This term represents a failure of DG13 to start. Phase II SAMAs C 5, 8, 9, 10, 14, and 16 for improving EDG reliability or adding additional onsite power sources were evaluated.

P81-MA-DGDG13-C 1.18E-02 1.0152 Diesel generator DG13 This term represents maintenance of DG13. Phase II SAMAs 5, unavailable due to 8, 9, 10, and 14 for improving EDG reliability or adding maintenance additional onsite power sources were evaluated.

E.1-19

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition R20-CF-CB-BKR 2.71E-07 1.0068 CCF of feeder breakers to This term represents a failure of DG11 to run. Phase II SAMAs LCCs feeding the chargers 5, 8, 9, 10, 14 and 15 for improving EDG reliability or adding 152-1604 & 1507 additional onsite power sources were evaluated.

R20-CF-TR15-16 9.02E-07 1.0507 CCF of LCC transformers for This term represents a failure of the LCC transformers for the the 15AA AND 16AB buses 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-B 8.40E-06 1.0053 Feeder breaker 152-1604 This term represents a failure of the power to LCC 16BB1 and fails open 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- 1.00E+00 1.0202 Operator fails to transfer to This term represents a failure of a human action to transfer to HEESFTRM alternate transformer 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-C 2.00E-03 1.0188 Fan cooler T51B001-C This term represents a failure of the HPCS pump room cooler.

unavailable due to Phase II SAMA 29 for adding HPCS HVAC procedures or maintenance hardware was evaluated.

X3 1.00E+00 1.0061 X3--depressurization via This term represents a failure to depressurize with RCIC during RCIC a SBO. Phase II SAMAs 1, 2, 3, 11, and 12 for adding or extending battery capacity were evaluated.

X77-FF-CFSTART- 4.85E-04 1.2754 X77 common cause start This term represents a failure of EDG area ventilation. Phase II U failures SAMAs 30, 32, and 33 for adding or enhancing EDG HVAC hardware were evaluated.

E.1-20

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition X77-FF-FSC001A-U 3.84E-03 1.0055 DG11 room vent start This term represents a failure of EDG area ventilation. Phase II failures SAMAs 30, 32, and 33 for adding or enhancing EDG HVAC hardware were evaluated.

X77-FF-FSC001B-U 3.84E-03 1.0053 DG12 room vent start This term represents a failure of EDG area ventilation. Phase II failures SAMAs 30, 32, and 33 for adding or enhancing EDG HVAC hardware were evaluated.

X77-FF-FSC002C-U 3.84E-03 1.0071 DIV 3 DG room vent start This term represents a failure of EDG area ventilation. Phase II faults SAMAs 30, 32, and 33 for adding or enhancing EDG HVAC hardware were evaluated.

Y47-FF-FSC01AA-U 6.43E-03 1.007 Y47 Train A start failures This 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-U 1.00E+00 1.0143 Operator fails to provide This term represents a failure of a human action to provide alternate cooling 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.

ZLLOCA 1.00E+00 1.036 Large LOCA sequence This term is a flag. No SAMAs need to be aligned.

ZS1LOCA 1.00E+00 1.024 Intermediate LOCA This term is a flag. No SAMAs need to be aligned.

sequences ZS2LOCA 1.00E+00 1.0084 Small LOCA sequences This term is a flag. No SAMAs need to be aligned.

ZSBO 1.00E+00 1.5093 SBO sequence (HPCS DG This term is a flag. No SAMAs need to be aligned.

fails)

E.1-21

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-2 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs (Based on CDF)

Event Name Probability RRW Event Description Disposition ZT1B 1.00E+00 1.0299 SBO sequence (HPCS DG This term is a flag. No SAMAs need to be aligned.

success)

ZTRAN 1.00E+00 2.2786 Transient sequence (no This term is a flag. No SAMAs need to be aligned.

SBO)

E.1-22

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage CDF Uncertainty The 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 95th 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.1.2 PSA Model - Level 2 Analysis E.1.2.1 Containment 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 (MAAP) 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 Table E.1-3.

E.1-23

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage The 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. Table E.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.

E.1-24

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-3 Notation and Definitions for GGNS CET Functional Nodes Description Parameter Monitored for CET Functional Node Success Criteria Success Determination Reactor Pressure Vessel This function questions whether the operator depressurizes the RPV after core RPV Pressure (< 100 psig)

(RPV) Depressurization damage but before vessel breach. Success of this action would allow low (OP) 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 alternativesc may be available but are not credited in this analysis.

E.1-25

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-3 (Continued)

Notation and Definitions for GGNS CET Functional Nodes Description Parameter Monitored for CET Functional Node Success Criteria Success Determination Core Melt Arrested In- In-vessel recovery or arrest of core melt progression addresses the ability of the < 1/2 core relocation Vessel (RX) operating staff to restore adequate core cooling from the time the end state of calculated by MAAP.

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 estimate success criteria 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 Guidelines (SAGs)) with capability greater than approximately 1000 gpm.d E.1-26

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-3 (Continued)

Notation and Definitions for GGNS CET Functional Nodes Description Parameter Monitored for CET Functional Node Success Criteria Success Determination Igniter Operation (H2) The functional success criterion for this event node is that the igniters operate The igniter hardware (1 of 2 as designed (1 of 2 divisions). divisions), the AC support system, the crew action to initiate, and the H2 analyzers allowing the initiation are all required for success.

Drywell (DW) Remains The functional success criteria for the DW intact node are that the DW retains its No energetic events and no Intact (CZ) pressure capability and that no early DW failure modes compromise the DW DW internal pressure integrity. The early DW failures modeled by the CZ node are characterized by > 65 psig.

phenomenological events (e.g., steam explosions, H2 deflagration, missile No energetic events and no generation, direct containment heating) that are estimated to challenge DW differential pressure containment integrity relatively quickly following core melt. Late DW failures,

> 42 psid.

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 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).

(cont. below)

E.1-27

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-3 (Continued)

Notation and Definitions for GGNS CET Functional Nodes Description Parameter Monitored for CET Functional Node Success Criteria Success Determination Drywell Remains Intact (CZ)

No in-vessel steam explosion (in-vessel steam explosions are precluded if (cont.) 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 The functional success criteria for the containment intact node are that the No energetic containment (CX) containment retains its pressure capability and that no early containment failure failure with internal pressure modes compromise the containment integrity. The early containment failures > 65 psig or the containment modeled by the CZ node are characterized by phenomenological events (e.g., profile curve.

steam explosions, H2 deflagration, missile generation, direct containment No containment differential heating) that are estimated to challenge containment integrity relatively quickly pressure > 42 psid.

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)

E.1-28

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-3 (Continued)

Notation and Definitions for GGNS CET Functional Nodes Description Parameter Monitored for CET Functional Node Success Criteria Success Determination Containment Remains Intact Note that successful prevention of early containment failure does not (CX) (cont.) 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.

E.1-29

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-3 (Continued)

Notation and Definitions for GGNS CET Functional Nodes Description Parameter Monitored for CET Functional Node Success Criteria Success Determination Containment Isolation (IS) The success of the containment isolation node (IS) is satisfied if the containment Failure size (< 2 inch dia.)

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.

Drywell Isolation (DL) The success of the drywell isolation node (DL) is that the drywell penetrations Failure size (< 2 inch dia.)

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.

E.1-30

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-3 (Continued)

Notation and Definitions for GGNS CET Functional Nodes Description Parameter Monitored for CET Functional Node Success Criteria Success Determination Debris Cooling (SI) Success at this node requires that water is available (greater than 1000 gpm) to Flow > 1000 gpm the core debris at the time of vessel failure or shortly thereafter (< 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ).

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 terminated 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.

E.1-31

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-3 (Continued)

Notation and Definitions for GGNS CET Functional Nodes Description Parameter Monitored for CET Functional Node Success Criteria Success Determination Containment Flooding Success at this node implies that the containment flooding contingency External flow > 1000 gpm Initiated (FC) procedure has been initiated by the operating staff and that a system of Vent > 6 inch dia.

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 conditions 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].

E.1-32

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-3 (Continued)

Notation and Definitions for GGNS CET Functional Nodes Description Parameter Monitored for CET Functional Node Success Criteria Success Determination Containment Pressure Successful containment pressure control is achieved if either of two functional 1. Cont. pressure < 65 psig Control nodes are successfully satisfied:

2. Cont. pressure < 22.4 psig (see node descriptions HR 1. Containment heat removal via pool cooling (Venting) and VC below) or

2. Containment venting Because these have different potential impacts on the radionuclide releases they are treated in separate nodes (see nodes HR and VC below).

E.1-33

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-3 (Continued)

Notation and Definitions for GGNS CET Functional Nodes Description Parameter Monitored for CET Functional Node Success Criteria Success Determination Containment Heat Removal Successful containment pressure control is unattainable using suppression pool Containment pressure < 65 (HR) cooling if either of the following conditions occurs: psig

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)

E.1-34

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-3 (Continued)

Notation and Definitions for GGNS CET Functional Nodes Description Parameter Monitored for CET Functional Node Success Criteria Success Determination Containment Venting (VC) The capability to vent the containment is a valuable supplement to the Containment pressure < 22.4 containment pressure control systems. As pressure and temperature increase, psig 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 6 inches 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.

E.1-35

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-3 (Continued)

Notation and Definitions for GGNS CET Functional Nodes Description Parameter Monitored for CET Functional Node Success Criteria Success Determination No Suppression Pool This node in the CET is used to characterize the magnitude of radionuclides that Bypass path < 6 inch dia.

Bypass (SP) 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.

Wetwell Airspace Breach This node appears after the No Suppression Pool Bypass node, i.e., drywell No WW water release path (WW) (Scrubbed Release) intact. This node distinguishes whether the wetwell failure occurred above or > 2 inch dia.

below the wetwell water line. Successfully avoiding a large containment failure Success (Up Branches) requires successful containment heat removal.

containment failure in the The probabilistic determination of the location of the failure is determined based dome (Wetwell Airspace.)

on the plant specific structural analysis for slow overpressurization events.

Containment Spray (CSS) This node distinguishes radionuclide release magnitude based on the 1 train of CSS operating availability of the CSS.

Enclosure Building/Auxiliary Preservation of the auxiliary building and enclosure building integrity results in a DF > 10 (Not currently Building Effective (EB) calculated decontamination factor (DF) using MAAP of > 10. 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 RPV level reaches top of active fuel. This was 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.

E.1-36

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage

d. 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 sufficient. 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).

E.1-37

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage H/E 1.05E-07 H/I 5% 1.23E-08 1%

H/L 8.73E-08 4%

M/E 3.49E-07 17%

Negligible (NCE) 8.73E-07 44%

M/I 1.73E-07 8%

LL/L M/L 6.83E-09 2.71E-07 0% 13%

LL/E 2.00E-09 L/E 0% 4.04E-09 0%

L/L LL/I L/I 1.32E-07 2.11E-09 3.34E-08 6%

0% 2%

Figure E.1-1 GGNS Radionuclide Release Category Summary Note: See Tables E.1-5 and E.1-6 for a definition of the release categories.

E.1-38

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage IIIB IIT 0%

1% IVA IIIA 0% IVL IIL 0%

IIIC 1%

0% 5%

IIA 8%

IA 38%

ID 11%

IBL 3%

IBE 33%

Figure E.1-2 Summary of GGNS Core Damage Accident Sequences Plant Damage States Note: Core Damage Accident Sequences Plant Damage State definitions can be seen in Table E.1-7.

E.1-39

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-4 Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)

Event Name Probability RRW Event Description Disposition B21-LF-FGCTISO 1.00E+00 1.0054 Containment isolation This term is a flag. No SAMAs need to be aligned.

signal present CX2-PH-CZF-NOTSU 5.46E-01 1.024 Containment success This term is a split fraction. No SAMAs need to be aligned.

during severe phenomena (CZ=F, CL II)

CX2-PH-CZS-NOTSU 9.82E-01 1.0173 Containment success This term is a split fraction. No SAMAs need to be aligned.

during severe phenomena (CZ=S, CL II)

CX--PH-CTCOND-F- 5.00E-01 6.317 Probability cont. fails This term represents a failure to control hydrogen or hydrogen given H2 late ignition ignition. Phase II SAMAs 44 and 45 for reducing the hydrogen detonation potential were evaluated.

CX--PH-H2-DEFGF- 1.00E+00 7.7756 Hydrogen This term represents a failure to control hydrogen or hydrogen deflagration occurs ignition. Phase II SAMAs 44 and 45 for reducing the hydrogen globally detonation potential were evaluated.

CX--PH-H2INVENF- 1.00E+00 7.7756 Sufficient hydrogen This term represents a failure to control hydrogen or hydrogen generated to cause ignition. Phase II SAMAs 44 and 45 for reducing the hydrogen overpressure detonation potential were evaluated.

CX-PH-LOOP-30MIN 8.00E-01 1.7986 AC power not This term represents a failure to recover offsite power. Phase recovered in 30 min 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-01 7.7756 Containment not This term represents a failure to control hydrogen or hydrogen inerted by steam ignition. Phase II SAMAs 44 and 45 for reducing the hydrogen detonation potential were evaluated.

E.1-40

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-4 (Continued)

Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)

Event Name Probability RRW Event Description Disposition CZ2-PH-ID-NOTSU 6.60E-01 1.0284 Drywell does not fail This term is a split fraction. No SAMAs need to be aligned.

due to severe phenomena (IGA=F, CLS ID)

CZ4-PH-IGF-NOTSU 9.08E-01 1.008 Drywell does not fail This term is a split fraction. No SAMAs need to be aligned.

due to severe phenomena (IGA=F)

CZ5-PH-IBE-NOTSU 9.13E-01 1.0871 Drywell does not fail This term is a split fraction. No SAMAs need to be aligned.

due to severe phenomena (CLASS IBE)

CZ--PH-2-NOTSU 9.87E-01 1.0177 Drywell does not fail This term is a split fraction. No SAMAs need to be aligned.

due to severe phenomena (CLASS II)

CZ--PH-CRDMELTF- 1.00E+00 1.1012 Control rods melt This term represents a possible reactivity excursion due to prior to fuel rods 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-01 5.3793 Conditional This term represents a failure to control hydrogen or hydrogen probability drywell ignition. Phase II SAMAs 44 and 45 for reducing the hydrogen fails given detonation potential were evaluated.

deflagration E.1-41

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-4 (Continued)

Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)

Event Name Probability RRW Event Description Disposition CZ--PH-FUELRODF- 1.00E-02 1.1012 Fuel rod integrity is This term represents timely restoration of emergency core maintined during the cooling to arrest the core melt progression in-vessel. Phase II reflood SAMAs 20, 21, 22, and 28 for improving high pressure injection capability were evaluated.

CZ--PH-SLCLWL-F- 1.00E+00 1.1012 Failure to inject SLC This term represents a failure of a human action to inject SLC with boron for low with boron for low water level. Phase II SAMAs 20 and 52 for water level improving high pressure injection and SLC capability were evaluated.

E12-FO-HECS-N 1.00E+00 1.0101 Operator fails to This term represents a failure of a human action to actuate actuate containment containment spray. Phase II SAMAs 46, 47, and 60 for spray improving containment vent capability were evaluated.

E12-FO-HEECCS-G 1.00E+00 1.0058 Operator fails to This term represents a failure of a human action to initiate low initiate LP ECCS pressure ECCS. Phase II SAMAs 20, 21, 22, and 28 for improving high pressure injection capability were evaluated.

E12-FO-HESPC-M 1.00E+00 1.0101 Operator fails to This term represents a failure of a human action to manually manually align for align for suppression pool cooling. Phase II SAMAs 46 and 47 suppression pool for improving containment vent capability were evaluated.

cooling E61-FO-H2-GB-X 1.00E+00 1.0074 Failure to obtain grab This term represents a failure of a human action obtain grab sample in SAPs sample in SAPs. Phase II SAMAs 44 and 45 for installing a passive hydrogen control system were evaluated.

E61-FO-IG-L1-X 1.00E+00 1.2301 Failure to initiate This term represents a failure of a human action to initiate igniters before igniters before transition to SAP. Phase II SAMAs 44 and 45 transition to SAP for installing a passive hydrogen control system were evaluated.

E.1-42

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-4 (Continued)

Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)

Event Name Probability RRW Event Description Disposition E61-FO-MSH13-X 1.00E+00 1.2191 Operator fails to This term represents a failure of a human action to energize energize hydrogen the hydrogen igniters. Phase II SAMAs 44 and 45 for installing igniters a passive hydrogen control system were evaluated.

EV 1.00E+00 1.0808 Early declaration of This term is a flag to represent an early declaration of a general emergency general emergency. No SAMAs need to be aligned.

G-IGNITION 5.38E-01 1.7464 Ignition source This term represents a failure to control hydrogen or hydrogen available at the ignition. Phase II SAMAs 44 and 45 for reducing the hydrogen incorrect time detonation potential were evaluated.

HI--PH-H2IGSBOF- 2.50E-01 1.7986 Random hydrogen This term represents a failure to control hydrogen or hydrogen ignition given no AC ignition. Phase II SAMAs 44 and 45 for reducing the hydrogen power detonation potential were evaluated.

IGA-PH-ID1-NOTSU 4.97E-01 1.0202 Igniters successful This term is a split fraction. No SAMAs need to be aligned.

(CLASS ID)

IGNITERS-FAIL 1.00E+00 1.0828 Igniters are operating This term is a flag. No SAMAs need to be aligned.

IGNITERS-SUC 1.00E+00 1.0356 Ingiters are operating This term is a flag. No SAMAs need to be aligned.

M41-FO-AVVCNT-Q 1.00E+00 1.0058 Operator fails to vent This term represents a failure of a human action to vent containment containment. Phase II SAMA 46 for a passive containment vent was evaluated.

NRC-L2-DEPB&IG 3.38E-05 1.0059 Failure to connect This term represents a failure of a human action to emergency ADS bottles and depressurize, igniter initiation in level 1, and igniter initiation in initiate H2 igniters level 2. Phase II SAMAs 44 and 45 for installing a passive hydrogen control system were evaluated.

E.1-43

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-4 (Continued)

Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)

Event Name Probability RRW Event Description Disposition NRC-OSP-DSG3 2.87E-02 1.0153 Fail to recover OSP This term represents a failure to recover offsite power. Phase given U2

3 FTR

II SAMAs 1, 2, 3, 11, 12 and 15 for extending available No SSW PHV failures recovery time by improving DC power were evaluated.

NRS-ALTPW&DEP 1.00E-06 1.0052 Failure to align This term represents a failure of a human action to align alternate power and alternate power and depressurize. Phase II SAMAs 1, 2, 3, 11, depressurize 12, and 15 for extending available recovery time by improving DC power were evaluated.

NRS-DHRLT 1.00E-07 1.0058 Failure to initiate SPC This term represents a failure of a human action to initiate SPC and containment and containment spray. Phase II SAMA 60 for improved spray containment heat removal were evaluated.

NRS-L2-DEP&IG 8.32E-06 1.0716 Failure to This term represents the operator to fail the following initiation:

depressurize and Emergency depressurization, igniter initiation in level 1, and start H2 igniters igniter initiation in level 2. Phase II SAMAs 44 and 45 for installing a passive hydrogen control system were evaluated.

NRS-L2-DEP&IG&FW 3.53E-06 1.0669 Failure to This term represents the operator to fail the following initiation:

depressurize and Emergency depressurization, igniter initiation in level 1, igniter start H2 igniters and initiation in level 2, and failure to restart FW. Phase II SAMAs restart FW pumps 44 and 45 for installing a passive hydrogen control system were evaluated.

NRS-L2-DEP&IG&PCS 1.43E-06 1.0243 Failure to This term represents the operator to fail the following initiation:

depressurize and Emergency depressurization, igniter initiation in level 1, igniter start H2 igniters and initiation in level 2, and align PCS. Phase II SAMAs 44 and 45 align PCS for installing a passive hydrogen control system were evaluated.

E.1-44

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-4 (Continued)

Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)

Event Name Probability RRW Event Description Disposition OP--AD-ALTRNT-F- 1.00E+00 1.0577 Alternate depress. This is a term to flag not crediting several primary system methods not credited depressurization schemes. No SAMAs need to be aligned.

OP--OP-DEPRESSH- 9.68E-01 1.0544 OP fails to depress This term represents a failure of a human action to given OP failed in depressurize given that the operator failed in the level 1 model LVL1 or loss of DC 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-NOTSU 7.11E-01 1.1331 Successful RPV This term is a split fraction. No SAMAs need to be aligned.

depressurization (Class IA, IE)

OP--PH-OP6-NOTSU 9.75E-01 1.0801 Successful RPV This term is a split fraction. No SAMAs need to be aligned.

depressurization (Class II)

OP--PH-PRESBK-F- 8.00E-01 1.0577 Pressure transient This term represents a high pressure vessel breach scenario does not fail where mechanical stress failures of the primary system mechanical systems pressure boundary failed to depressurize the RPV. There are no applicable SAMAs for this scenario.

OP--PH-SORV---F- 5.50E-01 1.0577 SRVs do not fail open This term represents a high pressure vessel breach scenario during core melt where the SRVs failed to stick open and allow progression depressurization. There are no applicable SAMAs for this scenario.

OP--PH-TEMPBK-F- 7.00E-01 1.0577 High prim sys temp This term represents a high pressure vessel breach scenario does not cause fail of where the RPV pressure boundary did not rupture due to high RCS press. bound internal RPV pressure and temperature. There are no applicable SAMAs for this scenario.

E.1-45

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-4 (Continued)

Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)

Event Name Probability RRW Event Description Disposition P41-CF-MV-DGIN-R 1.85E-04 1.0065 CCF of DG inlet isol This term represents a failure EDG cooling water due to MOVs FO18A-A AND isolation valve failures on the EDG. Phase II SAMAs 5, 8, 9, F018B-B to open and 10 for adding an additional generator and increasing the reliability of EDG cooling water were evaluated.

P64-PH-RX-EXO-F- 1.00E+00 1.2536 FPS (Paths 1-8) This term is a flag. No SAMAs need to be aligned.

inadequate for 1000gpm for Rx node P75-CF-DGR-Z 1.43E-03 1.0104 CCF of Div 1 & Div 2 This term represents a failure of the emergency AC power.

(& not Div 3) EDGs to Phase II SAMAs 5, 8, 9, 10, 11, 12 and 14 for improving EDG run reliability or adding additional onsite power sources were evaluated.

RP--OP-L2-CRODH- 1.00E+00 1.1012 Operator restores This term represents a failure of a human action to restore coolant injec. after ctrl coolant injection after the control rods are melted. Phase II rods are melted SAMAs 20, 21, 22, and 28 for improving high pressure injection capability were evaluated.

RXF 1.00E+00 1.0841 Failure of RX (OP=F This is a flag indicating that the RPV is at high pressure with or Classes IBE, II, low pressure injection systems not available or viable. No IIID, and IV) SAMAs need to be aligned.

RX--PH-RX2DNOTSU 1.09E-01 1.017 Core melt arrested in- This term is a split fraction. No SAMAs need to be aligned.

vessel (OP=S, Class ID)

RX--RX-FRECINJH- 9.00E-01 1.2536 Operator fails to This term represents a failure of a human action to recover recover injection injection before the RPV melt. Phase II SAMAs 20, 21, 22, before RPV melt and 28 for improving high pressure injection capability were evaluated.

E.1-46

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-4 (Continued)

Correlation of Level II Risk Significant Terms to Evaluated SAMAs (Based on Large Early Release Frequency)

Event Name Probability RRW Event Description Disposition SP--PH-BKFLOW-F- 1.00E-01 1.0352 No backflow if SPMU This term represents a suppression pool bypass after a core fails melt and vessel breach. Phase II SAMA 43 for installing a filtered vent was evaluated.

SP--PH-BKIGA-F- 1.00E+00 1.0826 No backflow if SPMU This term represents a suppression pool bypass after a core fails melt and vessel breach. Phase II SAMA 43 for installing a filtered vent was evaluated.

SP--VB-SEALS--F- 1.00E-02 1.0178 Temperature induced This term represents a suppression pool bypass after a core failure of all vacuum melt and vessel breach. Phase II SAMA 43 for installing a breaker seals filtered vent was evaluated.

SP--VB-SEALSNWF- 5.00E-02 1.0206 Temp induced failure This term represents a suppression pool bypass after a core of all vacuum breaker melt and vessel breach. Phase II SAMA 43 for installing a seals (RX=F, SI=F) filtered vent was evaluated.

WW--WW-L2-FAIL-- 1.00E-02 1.0196 Containment breach This term is a split fraction. No SAMAs need to be aligned.

below the wtr line (Class I, IIA, IIT, III, IV)

WW--WW-L2-NOT--- 9.90E-01 1.2389 Containment breach This term is a split fraction. No SAMAs need to be aligned.

above the wtr line (Class I, IIA, IIT, III, IV)

Note: 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.

E.1-47

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.1.2.2 Radionuclide Analysis E.1.2.2.1 Introduction A 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 considered, 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.1.2.2.2 Timing of Release Timing 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 0 hours 0 weeks 0 months 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 0 hours 0 weeks 0 months , but less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from declaration of a general emergency, for which offsite measures can be assumed to be fully effective.

E.1.2.2.3 Magnitude of Release Source 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 E.1-48

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage fraction 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 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 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 0.01 hours 5.952381e-5 weeks 1.3698e-5 months after accident initiation.

E.1.2.2.4 Release Category Bin Assignments Table E.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 Table E.1-6.

E.1-49

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-5 Release Severity and Timing Classification Scheme Summary Release Severity Release Timing Time of Initial Release Classification Cs Iodide % in Relative to Time for Classification Category Category Release General Emergency Declaration High (H) Greater than 10 Late (L) Greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Medium or 1 to 10 Intermediate (I) 4.0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Moderate (M)

Low (L) 0.1 to 1 Early (E) Less than 4.0 hours0 days 0 hours 0 weeks 0 months Low-low (LL) Less than 0.1 Negligible (NCF) 0 Table E.1-6 GGNS Release Categories Time of Magnitude of Release Release H M L LL E H/E M/E L/E LL/E I H/I M/I L/I LL/I L H/L M/L L/L LL/L E.1.2.2.5 Mapping of Level 1 Results into the Various Release Categories PDS 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.

E.1-50

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage From 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. Table E.1-7 provides a description of the GGNS PDS that are used to summarize the Level 1 results.

E.1-51

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-7 Summary of GGNS Core Damage Accident Sequences Plant Damage States Accident CAFTA Class Model Designator Subclass Definition (per Rx Yr)

Class I A Accident sequences involving loss of inventory makeup in which the reactor pressure 1.12E-06 remains high.

B Accident sequences involving a station blackout and loss of coolant inventory makeup. IBE = 9.71E-07 (Class IBE is defined as Early Station Blackout events with core damage at less than 4 IBL = 8.20E-08 hours. Class IBL is defined as Late Station Blackout events with core damage at greater than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .)

C Accident sequences involving a loss of coolant inventory induced by an ATWS sequence < 1E-12 with containment intact.

D Accident sequences involving a loss of coolant inventory makeup in which reactor pressure 3.17E-07 has been successfully reduced to 200 psi.

E Accident sequences involving loss of inventory makeup in which the reactor pressure (Grouped with remains high and DC power is unavailable. (Grouped with Class IA.) Class IA)

Class II A Accident sequences involving a loss of containment heat removal with the RPV initially 2.44E-07 intact; core damage; core damage induced post containment failure.

L Accident sequences involving a loss of containment heat removal with the RPV breached 9.84E-10 but no initial core damage; core damage induced post containment failure.

T Accident sequences involving a loss of containment heat removal with the RPV initially 1.37E-08 intact; core damage induced post high containment pressure.

V Class IIA and IIT except that the vent operates as designed; loss of makeup occurs at some < 1E-12 time following vent initiation. Suppression pool saturated but intact.

E.1-52

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-7 (Continued)

Summary of GGNS Core Damage Accident Sequences Plant Damage States Accident CAFTA Class Model Designator Subclass Definition (per Rx Yr)

Class III A Accident sequences leading to core damage conditions initiated by vessel rupture where 1.00E-08 (LOCA) the containment integrity is not breached in the initial time phase of the accident.

B Accident sequences initiated or resulting in small or medium LOCAs for which the reactor 6.39E-11 cannot be depressurized prior to core damage occurring.

C Accident sequences initiated or resulting in medium or large LOCAs for which the reactor is 1.60E-07 at low pressure and no effective injection is available.

D Accident sequences which are initiated by a LOCA or RPV failure and for which the vapor < 1E-12 suppression system is inadequate, challenging the containment integrity with subsequent failure of makeup systems.

Class IV A Accident sequences involving failure of adequate shutdown reactivity with the RPV initially 4.06E-09 (ATWS) intact; core damage induced post containment failure.

L Accident sequences involving a failure of adequate shutdown reactivity with the RPV initially (Grouped with breached (e.g., LOCA or stuck-open relief valve (SORV)); core damage induced post Class IVA) containment failure.

Class V --- Unisolated LOCA outside containment. 4.91E-10 Total CDF 2.92E-06 Note: 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.

E.1-53

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.1.2.2.6 Process Used to Group the Source Terms The 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 Table E.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 GGNS specific MAAP analyses are representative of each CET end state. It also bins each CET sequence into one of the release categories depicted in Table E.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 Table E.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.

E.1-54

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-8 Summary of Containment Event Tree Quantification Release Category Release Frequency (Magnitude/Timing) (Per ry)

H/E 1.05E-07 H/I 1.23E-08 H/L 8.73E-08 M/E 3.49E-07 M/I 1.73E-07 M/L 2.71E-07 L/E 4.04E-09 L/I 3.34E-08 L/L 1.32E-07 LL/E 2.00E-09 LL/I 2.11E-09 LL/L 6.83E-09 Negligible (NCF) 8.73E-07 CDF 2.05E-06 Nomenclature:

Timing (time between General Emergency Declaration and initial release):

Late (L) - Greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Intermediate (I) - 4.0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Early (E) - Less than 4.0 hours0 days 0 hours 0 weeks 0 months Magnitude:

Negligible (NCF)- Much less than 0.1% CsI release fraction Low-Low (LL) - Less than 0.1% CsI release fraction Low (L) - 0.1% to 1% CsI release fraction Medium (M) - 1% to 10% CsI release fraction High (H) - Greater than 10% CsI release fraction E.1-55

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.1.2.2.7 Consequence Analysis Source Terms Input 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 Table E.1-6. Table E.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.1.2.2.8 Release Magnitude Calculations The 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.

E.1-56

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-9 GGNS Release Category Source Terms Sheet 1 of 2 Release Release Mode Frequency Warning Time Elevation Release Start Duration Release (CET End State) (/year) (sec) (m) (sec) (sec) Energy (W)

H/E 1.05E-07 966 32 1497 257703 3.0E+07 H/I 1.23E-08 12 32 37707 221493 4.3E+05 H/L 8.73E-08 992 32 112096 111104 9.1E+04 M/E 3.49E-07 957 32 59664 199536 2.9E+06 M/I 1.73E-07 12 32 90866 168334 4.3E+05 M/L 2.71E-07 992 32 116944 142256 9.1E+04 L/E 4.04E-09 787 32 1279 4733 2.4E+06 L/I 3.34E-08 1264 32 1997 257203 4.3E+05 L/L 1.32E-07 966 32 107078 152122 8.3E+05 LL/E 2.00E-09 1266 32 1996 257204 2.9E+06 LL/I 2.11E-09 1265 32 1996 257204 4.3E+05 LL/L 6.83E-09 1266 32 186290 72910 9.1E+04 E.1-57

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-9 GGNS Release Category Source Terms Sheet 2 of 2 Release Fraction Release Mode (CET End State) NG I Cs Te Sr Ru La Ce Ba H/E 1.0E+00 1.4E-02 6.1E-03 5.5E-02 4.0E-04 6.5E-05 4.7E-05 5.0E-04 1.9E-04 H/I 1.0E+00 2.2E-01 7.6E-02 1.3E-01 9.8E-06 1.1E-05 5.4E-07 9.6E-06 1.0E-05 H/L 1.0E+00 1.7E-01 4.7E-02 1.5E-01 5.9E-06 5.7E-07 6.7E-07 7.3E-06 3.6E-06 M/E 8.8E-01 1.8E-01 5.2E-02 1.1E-01 1.5E-03 9.7E-04 1.4E-04 1.7E-03 1.2E-03 M/I 1.0E+00 3.6E-02 1.5E-01 1.1E-01 3.6E-06 1.1E-05 2.6E-07 4.5E-06 9.9E-06 M/L 1.0E+00 8.4E-02 5.0E-02 4.9E-02 2.2E-07 6.2E-07 1.6E-08 2.0E-07 1.3E-06 L/E 9.1E-01 2.1E-03 2.1E-03 2.1E-03 1.2E-05 3.9E-04 2.6E-07 1.5E-06 6.4E-05 L/I 1.0E+00 8.3E-02 2.5E-02 6.9E-02 1.7E-04 4.2E-05 4.4E-06 1.3E-04 9.2E-05 L/L 1.0E+00 7.2E-03 4.5E-03 4.3E-02 4.4E-06 1.4E-06 5.0E-07 6.1E-06 4.8E-06 LL/E 2.1E-02 5.3E-06 5.4E-07 1.9E-06 2.6E-09 2.3E-07 2.0E-10 1.1E-09 6.5E-08 LL/I 1.9E-02 1.7E-06 2.6E-07 3.0E-06 1.6E-09 2.2E-07 1.2E-10 6.6E-10 4.7E-08 LL/L 9.6E-01 1.0E-02 1.7E-02 1.0E-02 1.5E-06 1.8E-06 1.3E-07 1.2E-06 1.0E-06 E.1-58

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.1.3 IPEEE Analysis E.1.3.1 Seismic Analysis The 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.1.3.2 Fire 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].

Table E.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 experience 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.

E.1-59

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage

The peer review process for fire analyses was less well developed than for internal events PSAs. For example, no industry process, such as NEI 00-02, existed for the structured peer review of a fire PSA.

Plant-specific conservative assumptions 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.

E.1-60

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-10 GGNS Fire IPEEE Results Total Fire Compartment Compartment Description Compartment Screened(1)

CDF (/rx-yr)

CA101 Auxiliary Building Corridors. 93'-0" Elevation 5.74E-07 N CA102 RB Fire Zones 1A102, 1A202, 1A303, 1A442 1.01E-08 B CA103 RB Fire Zones 1A103, 1A203 4.85E-08 B CA104 RB Fire Zones 1A104, 1A204 6.76E-08 B CA105 RB Fire Zones 1A105, 1A205 7.19E-07 B CA106 RB Fire Zones 1A106, 1A206, 1A307, 1A441 8.56E-09 B CA107 RB Fire Zones 1A107 2.06E-08 A CA108 RB Fire Zones 1A108 2.06E-08 A CA109 RB Fire Zones 1A109 6.31E-07 B CA111 RB Fire Zones 1A111, 1A127 1.06E-08 A CA115 RB Fire Zones 1A115, 1A116, 1A118, 1A119, 1.34E-07 C 1A220 CA124 RB Fire Zones 1A124 7.35E-09 A CA125 RB Fire Zones 1A125 7.35E-09 A CA130 RB Fire Zones 1A130, 1A131 7.35E-09 A CA132 RB Fire Zones 1A132, 1A224, 1A226, 1A305, 8.83E-09 B 1A439, 1A440 CA201 Auxiliary Building Corridors. 119'-0" Elevation 6.38E-07 N CA207 Switchgear Room 1A207 3.47E-07 C CA208 Switchgear Room 1A208 8.14E-07 C CA209 RB Fire Zones 1A209 1.41E-08 A CA210 RB Fire Zones 1A210 1.41E-08 A CA219 Switchgear Room 1A219 4.09E-07 C CA221 Switchgear Room 1A221 4.57E-07 C CA225 RB Fire Zones 1A225 7.35E-09 A CA301 Auxiliary Building Corridors. 139'-0" Elevation 6.70E-07 N A422, 1A324 CA304 RB Fire Zones 1A304 7.36E-09 B CA306 RB Fire Zones 1A306 7.36E-09 B CA308 RB Fire Zones 1A308 2.62E-08 B CA309 RB Fire Zones 1A309 3.57E-07 B E.1-61

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-10 (Continued)

GGNS Fire IPEEE Results Total Fire Compartment Compartment Description Compartment Screened(1)

CDF (/rx-yr)

CA318 RB Fire Zones 1A318 2.41E-07 B CA319 RB Fire Zones 1A319 8.82E-09 A CA320 RB Fire Zones 1A320 3.09E-07 B CA323 RB Fire Zones 1A323 8.82E-09 A CA325 RB Fire Zones 1A325 7.35E-09 A CA326 RB Fire Zones 1A326 1.09E-08 A CA401 RB Fire Zones 1A401, 1A403, 1A417, 1A420, 1.94E-07 C 1A424, 1A427, 1A428, 1A434 CA402 RB Fire Zones 1A402 7.35E-09 A CA404 RB Fire Zones 1A404 7.35E-09 A CA405 RB Fire Zones 1A405 2.94E-08 A CA406 RB Fire Zones 1A406 1.12E-08 A CA407 Switchgear Room 1A407 5.00E-08 B CA410 Switchgear Room 1A410 5.00E-08 B CA429 RB Fire Zones 1A429 1.06E-08 A CA430 RB Fire Zones 1A430 2.73E-08 A CA431 RB Fire Zones 1A431, 1A437, 1A438, 1A444, 1.32E-07 A 1A525, 1A528, 1A532, 1A602, 1A603, 1A604, 1A606, 1A607 CA432 RB Fire Zones 1A432 2.38E-08 A CA433 RB Fire Zones 1A433 1.41E-08 A CA436 RB Fire Zones 1A436 7.35E-09 A CA506 RB Fire Zones 1A506, 1A508, 1A605 7.35E-09 A CA519 RB Fire Zones 1A519, 1A523, 1A524, 1A527, 2.26E-08 A 1A531 CA529 RB Fire Zones 1A529 2.73E-08 B CA530 RB Fire Zones 1A530 2.06E-08 A CA533 RB Fire Zones 1A533 7.35E-09 A CA534 RB Fire Zones 1A534 7.35E-09 A CA536 RB Fire Zones 1A536 7.35E-09 A CA537 RB Fire Zones 1A537 7.35E-09 A E.1-62

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-10 (Continued)

GGNS Fire IPEEE Results Total Fire Compartment Compartment Description Compartment Screened(1)

CDF (/rx-yr)

CA539 RB Fire Zones 1A539 3.45E-07 C CC101 RB Fire Zones OC101, OC103, OC115, 5.67E-08 C OC117, OC217 CC104 Hot Machine Shop 2.42E-07 N CC125 RB Fire Zones OC125 1.18E-08 A CC126 RB Fire Zones OC126 2.06E-08 A CC128 RB Fire Zones OC128 1.38E-08 B CC202 Division 1 Switchgear Room 9.37E-07 N CC203 RB Fire Zones OC203 2.94E-07 C CC204 RB Fire Zones OC204 1.35E-08 A CC205 RB Fire Zones OC205 7.35E-09 A CC205A RB Fire Zones OC205A 7.35E-09 A CC206 RB Fire Zones OC206 7.35E-09 A CC207 Battery Room OC207 8.84E-07 C CC208 RB Fire Zones OC208 8.75E-08 B CC208A RB Fire Zones OC208A 3.75E-08 B CC209 Battery Room OC209 4.63E-07 B CC210 Division 3 (HPCS) Switchgear Room 6.08E-07 N CC211 Battery Room OC211 2.94E-07 C CC212 RB Fire Zones OC212 7.35E-09 A CC213 RB Fire Zones OC213 3.86E-08 B CC214 RB Fire Zones OC214 4.24E-07 C CC215 Division 2 Switchgear Room 4.06E-07 N CC216 RB Fire Zones OC216 7.36E-09 B CC218 RB Fire Zones OC218 7.35E-09 A CC219 RB Fire Zones OC219 7.35E-09 A CC301 RB Fire Zones OC301 7.35E-09 A CC302 HVAC Equipment Room 2.10E-07 N CC303 RB Fire Zones OC303 4.42E-08 B CC304 RB Fire Zones OC304, OC412, OC612 4.46E-08 B E.1-63

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-10 (Continued)

GGNS Fire IPEEE Results Total Fire Compartment Compartment Description Compartment Screened(1)

CDF (/rx-yr)

CC305 RB Fire Zones OC305 7.35E-09 A CC306 RB Fire Zones OC306, OC409, OC610, 1.03E-08 A OC709 CC307 RB Fire Zones OC307 1.94E-07 C CC308 RB Fire Zones OC308 7.36E-09 B CC309 RB Fire Zones OC309 7.35E-09 A CC401 RB Fire Zones OC401 7.35E-09 A CC402 Cable Spreading Room 2.82E-07 N CC402A RB Fire Zones OC402A, OC512B 7.35E-09 A CC403 RB Fire Zones OC403 1.04E-07 C CC404 RB Fire Zones OC404 7.35E-09 A CC405 RB Fire Zones OC405 7.35E-09 A CC405A RB Fire Zones OC405A, OC507A 7.35E-09 A CC406 RB Fire Zones OC406 7.35E-09 A CC406A RB Fire Zones OC406A, OC518A, OC613A 7.35E-09 A CC407 RB Fire Zones OC407 1.35E-07 A CC408 RB Fire Zones OC408 1.05E-07 B CC409A RB Fire Zones OC409A, OC512, OC608B 7.35E-09 A CC410 Battery Room OC410 1.91E-08 A CC411 RB Fire Zones OC411 7.35E-09 A CC412A RB Fire Zones OC412A, OC507C, OC603B 7.35E-09 A CC501 RB Fire Zones OC501 7.35E-09 A CC502 Control Room 3.85E-06 N CC507 RB Fire Zones OC507 7.35E-09 A CC509 RB Fire Zones OC509, OC511, OC512 7.35E-09 A CC510 RB Fire Zones OC510 7.35E-09 A CC513 RB Fire Zones OC513 7.35E-09 A CC514 RB Fire Zones OC514 7.35E-09 A CC515 RB Fire Zones OC515 7.35E-09 A CC518 RB Fire Zones OC518, OC611 7.35E-09 A E.1-64

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-10 (Continued)

GGNS Fire IPEEE Results Total Fire Compartment Compartment Description Compartment Screened(1)

CDF (/rx-yr)

CC601 RB Fire Zones OC601, OC602 4.58E-08 B CC603 RB Fire Zones OC603 1.12E-08 A CC604 RB Fire Zones OC604 9.70E-09 A CC606 RB Fire Zones OC606 7.35E-09 A CC608 RB Fire Zones OC608 1.03E-08 A CC609 RB Fire Zones OC609 7.35E-09 A CC613 RB Fire Zones OC613 7.35E-09 A CC614 RB Fire Zones OC614 9.11E-09 A CC615 RB Fire Zones OC615 7.35E-09 A CC616 RB Fire Zones OC616 7.35E-09 A CC617 RB Fire Zones OC617 7.35E-09 A CC618 RB Fire Zones OC618 1.05E-07 B CC619 RB Fire Zones OC619 7.35E-09 A CC701 RB Fire Zones OC701 7.35E-09 A CC702 Cable Spreading Room OC702 5.18E-07 C CC703 RB Fire Zones OC703 4.72E-07 C CC704 RB Fire Zones OC704 2.12E-08 A CC705 RB Fire Zones OC705 8.82E-09 A CC706 RB Fire Zones OC706 1.05E-07 B CC707 RB Fire Zones OC707 1.09E-07 A CC708 RB Fire Zones OC708, OC710 5.59E-08 A CC708A RB Fire Zones OC708A 7.35E-09 A CC711 RB Fire Zones OC711 7.35E-09 A CC712 RB Fire Zones OC712 7.35E-09 A CC713 RB Fire Zones OC713 7.35E-09 A CD301 RB Fire Zones 1D301 4.34E-08 B CD306 Division 3 (HPCS) Diesel Generator Room 1.72E-07 N CD308 Diesel Generator Room ID308 3.05E-07 C CD310 Diesel Generator Room ID310 3.48E-07 C CM100 RB Fire Zones BASIN NO. 1 2.80E-08 B E.1-65

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-10 (Continued)

GGNS Fire IPEEE Results Total Fire Compartment Compartment Description Compartment Screened(1)

CDF (/rx-yr)

CM110 RB Fire Zones IM110 2.56E-07 C CM112 RB Fire Zones IM112 7.90E-07 B CM200 RB Fire Zones BASIN NO. 2 2.90E-08 B CM210 RB Fire Zones 2M110 7.31E-07 B CM212 RB Fire Zones 2M112 4.02E-08 B CT100 Turbine Building Floor, 93'-0" Elevation 3.24E-07 N CT200 Turbine Building Floor, 113'-0" Elevation 7.10E-09 N CT212 Battery Room 1T212 1.91E-08 A CT219 Switchgear Room 1T219 9.23E-07 B CT300 133'-0" Elevation, Turbine Bldg. 5.19E-07 C CT312 Battery Room 1T312 1.91E-08 A CT323 Switchgear Room 1T323 8.84E-07 B CT400 166'-0" Elevation, Turbine Bldg. + 1T502, 1.63E-07 C 1T503 CT405 Battery Room 1T405 1.91E-08 A CT406 Battery Room 1T406 1.91E-08 A CM101 OM101 (Circ. Water Pumphouse) 6.48E-08 B CM102 OM102 (Mtr. Driver Fire Pump Room) 1.14E-07 B CM115 OM115 (all Water Treatment Bldg.) 2.42E-07 C CRAD Radwaste Bldg. 3.29E-07 C CTR11 Transformers BOP11A, BOP11B 1.94E-07 C CTR12 Transformers BOP12A, BOP12B 1.65E-07 B CTR14 Transformers BOP14, BOP24 4.70E-08 A CTRMAIN Transformers Main 1A, 1B, 1C, 1D 8.53E-08 A CDUC1 Division 1 duct bank to SSW Cooling Tower 3.15E-08 B CDUC2 Division 2 duct bank to SSW Cooling Tower 2.47E-07 C CDUC3 Division 3 duct bank to SSW Cooling Tower 2.52E-07 B YARD Balance of Yard Area 7.71E-07 C Total 2.74E-05

Reference:

E.1-15 E.1-66

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage

1. Screening Criteria in Table E.1-10:

A Screened based on no safe shutdown or PRA equipment.

B Screened assuming all equipment in compartment is failed.

C Screened with credit for detailed recovery.

N Not screened, more detailed analysis performed.

E.1.3.3 Other External Hazards The 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.

E.1-67

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.1.4 PSA Model Revisions and Peer Review Summary The summary of the GGNS PSA models CDF and LERF is presented in the table below.

Summary of Major PSA Models PSA Model CDF (/rx-yr) LERF (/rx-yr) 1992 (IPE) 1.72E-05 5.17E-07 1997 (R1) 5.46E-06 Not Updated 2002 (R2) 4.27E-06 2.04E-07 2010 (R3) 2.69E-06 1.44E-07 2010 (EPU) 2.91E-06 1.48E-07 E.1.4.1 Major Differences between the 1997(R1) PSA Model and the IPE Model The 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 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.1.4.2 Major Differences between the 2002 (R2) PSA Model and the 1997(R1) PSA Model The 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.

E.1-68

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.1.4.3 Major Differences between the 2010 (R3) PSA Model and the 2002 (R2) PSA Model The 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 CRDless 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.1.4.4 Major Differences between the 2010 (EPU) PSA Model and the 2010 (R3) PSA Model The 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 E.1-69

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage

Hardware changes

Procedural changes

Operational changes The 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 Models Contributing Initiator R1 R2 R3 R3 EPU Group LOSP 42.5% 38.6% 38.7% 39.5%

Loss of Feedwater (FW) 4.3% 21.2% 8.6% 8.1%

PCS Avail Trans 5.9% 16.1% 20.8% 20.5%

Loss of PCS 7.7% 12.1% 12.9% 12.4%

Special(1) 30.6% 11.3% 7.9% 7.8%

LOCA 7.9% 0.4% 4.0% 3.8%

SORV 0.1% 0.1% 0.3% 0.3%

ISLOCA not modeled 0.1% 0.0% 0.0%

Flood 1.0% 0.1% 0.1% 0.0%

E.1-70

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Contribution to CDF Changes in PRA Models (Continued)

Contributing Initiator R1 R2 R3 R3 EPU Group RPV Rupture not modeled not modeled 0.4% 0.3%

Loss of Service not modeled not modeled 6.5% 7.1%

Transformer

1. Special initiators include loss of AC bus, DC bus, service water, closed cooling water, or instrument air.

E.1.4.5 PSA Model Peer Review The 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.1.5 The MACCS2 ModelLevel 3 Analysis E.1.5.1 Introduction 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 E.1-71

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Consequences 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.1.5.2 Input The following sections describe the site-specific input parameters used to obtain the off-site dose and economic impacts for cost-benefit analyses.

E.1.5.2.1 Projected Total Population by Spatial Element The 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 Table E.1-11.

E.1-72

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-11 Estimated Population Distribution within a 50-Mile Radius Wind 0 to 11 to 21 to 31 to 41 to Total Direction 10 miles 20 miles 30 miles 40 miles 50 miles N 25 91 659 765 665 2,205 NNE 27 5,447 34,095 4,354 797 44,720 NE 122 1,938 4,538 5,201 3,913 15,712 ENE 252 239 3,495 4,784 70,710 79,480 E 404 656 1,561 4,750 16,603 23,974 ESE 1,320 1,043 354 6,931 10,849 20,497 SE 3,436 1,371 739 3,407 28,418 37,371 SSE 602 1,158 584 3,472 2,668 8,484 S 124 2,353 4,881 1,949 1,383 10,690 SSW 736 1,426 2,445 29,732 4,606 38,945 SW 250 375 1,493 14,646 3,387 20,151 WSW 88 1,740 297 1,781 2,162 6,068 W 103 316 351 4,505 3,080 8,355 WNW 20 2,409 263 4,451 11,260 18,403 NW 12 136 57 2,239 6,332 8,776 NNW 3 94 11,567 714 2,830 15,208 Totals 7,524 20,792 67,379 93,681 169,663 359,039 E.1-73

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.1.5.2.2 Land Fraction The 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.1.5.2.3 Watershed Class Watershed 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.1.5.2.4 Regional Economic Data Region Index Each 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 Data Economic 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 Wealth MACCS2 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 Wealth MACCS2 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.

E.1-74

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Variable Description Value CHEVACST001 Daily cost for a person who has been evacuated ($/person- 51.3 day)

CHPOPCST001 Population relocation cost ($/person) 9500 CHRELCST001 Daily cost for a person who is relocated ($/person-day) 51.3 CHCDFRM001 Cost of farm decontamination for the various levels of 1068.75 decontamination ($/hectare) 2375 CHCDNFRM001 Cost of non-farm decontamination for the various levels of 5700 decontamination ($/person) 15200 CHDLBCST001 Average cost of decontamination labor ($/person-year) 66500 DPRATE Property depreciation rate (per year) 0.2 DSRATE Investment rate of return (per year) 0.12 E.1.5.2.5 Agriculture Data The 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.1.5.2.6 Meteorological Data The 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 Data Meteorological 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 Data Mixing 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.

E.1-75

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.1.5.2.7 Emergency Response Assumptions A 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 Time The 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 Speed The 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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.1.5.2.8 Core Inventory The GGNS core inventory is shown in Table E.1-12. These values are based on ORIGEN 2.1 evaluations supporting the EPU to 115% (4408 MWt) of the original licensed thermal power.

E.1-76

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-12 Estimated GGNS Core Inventory (Becquerels)(1)

Nuclide Inventory Nuclide Inventory Co-58 4.22E+16 Te-131m 6.59E+17 Co-60 7.29E+16 Te-132 6.40E+18 Kr-85 5.81E+16 I-131 4.51E+18 Kr-85m 1.19E+18 I-132 6.51E+18 Kr-87 2.28E+18 I-133 9.18E+18 Kr-88 3.20E+18 I-134 1.01E+19 Rb-86 1.12E+16 I-135 8.58E+18 Sr-89 4.33E+18 Xe-133 8.81E+18 Sr-90 4.63E+17 Xe-135 3.12E+18 Sr-91 5.40E+18 Cs-134 1.04E+18 Sr-92 5.85E+18 Cs-136 3.35E+17 Y-90 4.92E+17 Cs-137 6.18E+17 Y-91 5.59E+18 Ba-139 8.18E+18 Y-92 5.88E+18 Ba-140 7.92E+18 Y-93 6.77E+18 La-140 8.40E+18 Zr-95 7.99E+18 La-141 7.47E+18 Zr-97 8.25E+18 La-142 7.22E+18 Nb-95 8.03E+18 Ce-141 7.51E+18 Mo-99 8.55E+18 Ce-143 6.96E+18 Tc-99m 7.44E+18 Ce-144 6.14E+18 Ru-103 7.14E+18 Pr-143 6.73E+18 Ru-105 5.03E+18 Nd-147 3.00E+18 Ru-106 2.76E+18 Np-239 9.32E+19 Rh-105 4.74E+18 Pu-238 1.89E+16 Sb-127 5.00E+17 Pu-239 1.91E+15 Sb-129 1.48E+18 Pu-240 2.58E+15 E.1-77

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-12 (Continued)

Estimated GGNS Core Inventory (Becquerels)(1)

Nuclide Inventory Nuclide Inventory Te-127 4.96E+17 Pu-241 8.44E+17 Te-127m 6.70E+16 Am-241 9.44E+14 Te-129 1.45E+18 Cm-242 2.50E+17 Te-129m 2.16E+17 Cm-244 1.58E+16

1. From GGNS specific data for a power level of 4408 MWth [E.1-2].

E.1.5.2.9 Source Terms Eleven 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.1.5.3 RESULTS 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 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months (base) to 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . Sensitivity Case 2 assumes a lower average evacuation speed; the speed was reduced from 4.87 m/s (base) to 2.435 m/s.

Table E.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.

E.1-78

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-13 Base Case Mean PDR and OECR Values for Postulated Internal Events Offsite Offsite Population Economic Characteristics of Economic Dose Risk Cost Risk Release Mode(1) Population Dose Cost (PDR) (OECR)

Frequency (person- (person-ID person-rem ($) $/yr (per year) sv)(1) rem/yr)(2)

H/L 8.73E-08 4.12E+03 4.12E+05 1.04E+09 3.60E-02 9.08E+01 H/E 1.05E-07 2.29E+03 2.29E+05 3.13E+08 2.41E-02 3.29E+01 H/I 1.23E-08 5.10E+03 5.10E+05 1.37E+09 6.26E-03 1.68E+01 M/E 3.49E-07 4.66E+03 4.66E+05 1.31E+09 1.63E-01 4.57E+02 M/I 1.73E-07 6.70E+03 6.70E+05 1.81E+09 1.16E-01 3.14E+02 M/L 2.71E-07 3.86E+03 3.86E+05 1.04E+09 1.05E-01 2.82E+02 L/E 4.04E-09 9.92E+02 9.92E+04 7.32E+07 4.00E-04 2.95E-01 L/I 3.34E-08 3.26E+03 3.26E+05 7.48E+08 1.09E-02 2.50E+01 L/L 1.32E-07 1.75E+03 1.75E+05 1.66E+08 2.30E-02 2.18E+01 LL/E 2.00E-09 3.62E+00 3.62E+02 4.63E+05 7.24E-07 9.26E-04 LL/I 2.11E-09 1.80E+00 1.80E+02 4.59E+05 3.80E-07 9.68E-04 LL/L 6.83E-09 2.90E+03 2.90E+05 4.81E+08 1.98E-03 3.28E+00 Totals 4.86E-01 1.24E+03

1. Conversion Factor: 1 sv = 100 rem.

2. Value is the product of the release mode frequency and the population dose.

Results 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. Table E.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.

E.1-79

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.1-14 Summary of Offsite Consequence Results for Sensitivity Results Population Dose (person-sv)(1) Offsite Economic Cost ($)

Longer Time Slower Longer Time Slower Release Base for Speed of Base for Speed of Mode Evacuation Evacuation Evacuation Evacuation H/L 4.12E+03 4.12E+03 4.12E+03 1.04E+09 1.04E+09 1.04E+09 H/E 2.29E+03 2.30E+03 2.29E+03 3.13E+08 3.13E+08 3.13E+08 H/I 5.10E+03 5.10E+03 5.10E+03 1.37E+09 1.37E+09 1.37E+09 M/E 4.66E+03 4.66E+03 4.66E+03 1.31E+09 1.31E+09 1.31E+09 M/I 6.70E+03 6.70E+03 6.70E+03 1.81E+09 1.81E+09 1.81E+09 M/L 3.86E+03 3.86E+03 3.86E+03 1.04E+09 1.04E+09 1.04E+09 L/E 9.92E+02 9.93E+02 9.96E+02 7.32E+07 7.32E+07 7.32E+07 L/I 3.26E+03 3.27E+03 3.26E+03 7.48E+08 7.48E+08 7.48E+08 L/L 1.75E+03 1.75E+03 1.75E+03 1.66E+08 1.66E+08 1.66E+08 LL/E 3.62E+00 3.65E+00 3.63E+00 4.63E+05 4.63E+05 4.63E+05 LL/I 1.80E+00 1.83E+00 1.80E+00 4.59E+05 4.59E+05 4.59E+05 LL/L 2.90E+03 2.90E+03 2.90E+03 4.81E+08 4.81E+08 4.81E+08 Total 4.86E-01 4.86E-01 4.86E-01 1.24E+03 1.24E+03 1.24E+03 person-rem/ person-rem/ person-rem/ $/yr $/yr $/yr yr yr yr

1. Conversion Factor: 1 sv = 100 rem.

E.1-80

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.1.6 References E.1-1 NEI 05-01, Severe Accident Mitigation Alternatives (SAMA) Analysis Guidance Document, November 2005, Revision A.

E.1-2 CALC-OC-N1000-10002, "GGNS Level 3 Probabilistic Safety Analysis (PSA) Model" Rev. 0.

E.1-3 Grand Gulf Nuclear Station Development of Evacuation Time Estimates, KLD Associates, Inc., September 2007.

E.1-4 PRA-GG-01-003, Grand Gulf Power Station Detailed Level 2 Analysis, Revision 0, August 2010.

E.1-5 GGNS PRA-GG-01-001, "GGNS Level-1 Model Revision 3 PSA Summary Report,"

Rev. 2.

E.1-6 Intentionally Left Blank E.1-7 Kaiser, "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.1-8 USNRC, NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, 1991.

E.1-9 GGNS94-0053 IPEEE, "Internal Plant Examination of External Events Seismic Margins," Revision 0.

E.1-10 USNRC, NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, June 1991.

E.1-11 EPRI NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," August 1991.

E.1-12 EPRI 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 Response to Generic Letter 88-20, 'Individual Plant Examination for Severe Accidents Vulnerabilities,'" Correspondence No. GNRO-92/

00157, letter dated December 23, 1992.

E.1-14 GGNS-97-0014, "GGNS PRA Update Summary and Results Report," July 30, 1997.

E.1-15 GGNS95-00041, "Internal Plant Examination of External Events Fire," October 1996.

E.1-81

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.1-16 USNRC 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.1-17 GGNS Calculation No. XC-N1111-01007, "GGNS Level 1 PSA," Revision 2, October 17, 2002.

E.1-18 GGNS Calculation No. PRA-GG-09-001, "Identification of Risk Implications due to Extended Power Uprate at Grand Gulf," May 2010.

E.1-82

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage ATTACHMENT E.2 EVALUATION OF GGNS SAMA CANDIDATES

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.2 EVALUATION OF GGNS SAMA CANDIDATES This section describes the generation of the initial list of potential SAMA candidates, screening methods, and the analysis of the remaining SAMA candidates.

E.2.1 SAMA List Compilation Candidate 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 Individual Plant Examination 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 Table E.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 similar 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].

E.2-1

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.2.2 Qualitative 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. Table E.2-2 provides a description of each of the 63 Phase II SAMA candidates.

E.2.3 Final 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.

E.2-2

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage As 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.

Type of Change Estimated Cost Range Procedural only $25K-$50K Procedural change with engineering or training $50K-$200K required Procedural change with engineering and testing/ $200K-$300K training required Hardware modification $100K to > $1000K 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 functions 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 Table E.2-2.

Bounding evaluations (or analysis cases) were performed to address specific 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 Power This 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 E.2-3

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage the 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 Reliability This 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 level 1 PSA model. The following basic events were removed from the model:

11DA-007-D 11DA-008-D 11DB-007-E 11DB-008-E 11DC-007-F 11DC-008-F 11DD-007-X 11DD-008-X 11DE-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-X L21-CO-CB11E02-X L21-CO-CB11E03-X L51-LP-BC-1A4-D L51-LP-BC-1A5-D L51-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-D L51-MA-BC-1B4-E L51-MA-BC-1B5-E L51-MA-BC-1D4-X L51-MA-BC-1D5-X L51-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-F P81-MA-BC-1C4-F P81-MA-BC-1C5-F R20-CF-CB-BKR R20-CO-CB15102-X R20-CO-CB15202-X R20-CO-CB15306-D R20-CO-CB15602-D R20-CO-CB16102-X R20-CO-CB16202-X R20-CO-CB16306-E R20-CO-CB16602-E R20-CO-CB31116-F 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-Ties This analysis case was used to evaluate the change in plant risk from providing DC bus cross-ties. A bounding analysis was performed 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.

E.2-4

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Case 4: Increase Availability of On-Site AC Power This 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 Power This 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 Weather This 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) Cooling This 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 Reliability This 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-E.2-5

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage DG12-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 reliability This 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 Power This 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 model. 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) Valves This 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:

15P21-001 PROB 0

16P41-001 PROB 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 System This 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 E.2-6

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage uncertainty) 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 System This 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 Signals This 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 basic 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 System This 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 Interlock This 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 E.2-7

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage and 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 Exchangers This 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 Reliability This 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 Reliability This 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 Cooling This 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 Improvements This 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 E.2-8

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage three 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 Cooling This 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 Air This 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 set 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 SRV This 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 MSIVs This 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 Cooling This 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 E.2-9

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage the failure of flow to the RHR heat exchangers in the PSA model [gates P41-RHRHXA, P41-RHRHXB, P41-RHRHXA-SBO, and P41-RHRHXB-SBO were 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 Removal This 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 CapabilityDrywell Spray This 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 CST This 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.

E.2-10

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Case 32: Reduce Hydrogen Ignition This 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 Venting This 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 Design This 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) System This 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.

E.2-11

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Case 37: SRV Reseat This 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 SRVs 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 Suppression This 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 CO2 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 CO2 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 CO2 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

[Table E.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.

E.2-12

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage 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 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 [Table 4.21-1]

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

Internal events CDF = 2.05E-06/rx-yr Maximum fire benefit = Maximum internal benefit x Total fire CDF/Internal events CDF Maximum fire benefit = $74,673 x (2.74E-05/2.05E-06)= $997,559 SAMA case 38 benefit = 3.42% x (Maximum fire benefit) = 0.0342 x $997,559 SAMA case 38 benefit = $34,115 Applying the uncertainty factor of 3, SAMA case 38 benefit with uncertainty = $34,115 x 3 = $102,345 Case 39: Reduce Risk from Fires that Require Control Room Evacuation The 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.

E.2-13

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage

Given, Maximum internal benefit is $74,673 [Table 4.21-1]

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

Internal events CDF = 2.05E-06/rx-yr Maximum fire benefit = Maximum internal benefit x Total fire CDF/Internal events CDF Maximum fire benefit = $74,673 x (2.74E-05/2.05E-06) = $997,599 SAMA case 39 benefit = 14.05% x (Maximum fire benefit) = 0.1405 x $997,599 SAMA case 39 benefit = $140,174 Applying the uncertainty factor of 3, SAMA case 39 benefit with uncertainty = $140174 x 3 = $420,521 Case 40: Large Break LOCA This 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 Risk This 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 System This 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.

E.2-14

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Case 43: Increase Recovery Time of ECCS upon Loss of SSW This 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 Removal This 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 Availability This 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 Cooling This 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.2.4 Sensitivity Analyses Two sensitivity analyses were conducted to gauge the impact of assumptions upon the analysis.

The benefits estimated for each of these sensitivities are presented in Table E.2-3.

A description of each sensitivity case follows.

E.2-15

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Sensitivity Case 1: Years Remaining Until End of Plant Life The 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 Rate The 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.

E.2-16

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.2.5 References E.2-1 Nuclear Energy Institute (NEI), NEI 05-01, Severe Accident Mitigation Alternatives (SAMA) Analysis Guidance Document, November 2005, Revision A.

E.2-2 U.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.2-3 USNRC, 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.2-4 USNRC, 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.2-5 USNRC, 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.2-6 USNRC, 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.2-7 USNRC, 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.2-8 USNRC, NUREG-1742 Perspectives Gained From the Individual Plant Examination of External Events (IPEEE) Program, Volumes 1 & 2, Final Report April 2002.

E.2-9 USNRC, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Cooper Nuclear Station (NUREG-1437, Supplement 41)

Final Report, July 2010.

E.2-10 USNRC, 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.2-11 USNRC, 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.

E.2-17

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage E.2-12 USNRC, 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.2-13 GGNS94-0054 IPEEE "Internal Plant Examination of External Events Summary Report,"

Revision 1.

E.2-14 GGNS95-00041 IPEEE "Internal Plant Examination of External Events Fire," Revision 0.

E.2-15 GGNS94-0053 IPEEE "Internal Plant Examination of External Events Seismic Margins,"

Revision 0.

E.2-16 GGNS94-0051IPEEE "Internal Plant Examination of External Events Fire Modeling,"

Revision 1.

E.2-17 GGNS93-0048 IPEEE "Internal Plant Examination of External Events High Wind and Tornado Assessment," Revision 0.

E.2-18 Grand Gulf Nuclear Station Individual Plant Examination Summary Report (IPE),

December 1992.

E.2-19 Energy Northwest, License Renewal Application, Appendix E, Applicants Environmental Report, Operating License Renewal Stage, Columbia Generating Station, Energy Northwest," January 2010.

E.2-20 USNRC, NUREG/BR-0184, Regulatory Analysis Technical Evaluation Handbook, January 1997.

E.2-21 CALC-OC-N1000-10003, Cost-Benefit Analysis of Severe Accident Mitigation Alternatives, Revision 0.

E.2-18

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-1 Phase I SAMAs Related to IPE and IPEEE Insights Phase I Credited SAMA Result of Potential Screening SAMA Title SAMA Disposition in PSA ID Enhancement Results Model Number 226 The Loss of Offsite Power Off- Increased availability of #3 - Already The Loss of AC Power Off-Normal Yes Normal Event Procedure will be on-site AC power leading installed Event Procedure has been revised revised to allow for the Level 2 to increased availability of to allow for the level 2 signal to be signal to be bypassed in the event ECCS injection. bypassed in the event that the that the Division 3 diesel generator Division 3 diesel generator must be must be cross-tied to Divisions 1 or cross-tied to divisions 1 or 2.

2.

E.2-19

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-1 (Continued)

Phase I SAMAs Related to IPE and IPEEE Insights Phase I Credited SAMA Result of Potential Screening SAMA Title SAMA Disposition in PSA ID Enhancement Results Model Number 227 Improve secondary containment Improved availability of #3 - Already The PSW isolation valves in the Yes isolation to allow the capability of PSW and Instrument Air installed Auxiliary Building penetrations bypassing the isolation signals and such that the main (P44-F1l6, P44-FI17, P44-FII8, re-opening the valves. condenser, condensate, P44-FII9, P44-FI20, P44-FI2I, P44-and feedwater systems FI22 and P44-FI23) can be would not be lost. CRD reopened by manual override after would also not be a LOCA to reestablish PSW cooling degraded due to a loss of to the CCW heat exchangers, the preferred cooling computer room coolers, plant source of the component chillers, steam tunnel coolers, and cooling water (CCW) heat drywell coolers. This should be exchangers. 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.

E.2-20

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-1 (Continued)

Phase I SAMAs Related to IPE and IPEEE Insights Phase I Credited SAMA Result of Potential Screening SAMA Title SAMA Disposition in PSA ID Enhancement Results Model Number 228 Implement procedural changes to Increased RCIC #1 - N / A Provided there is no leak in the No allow for bypass of the RCIC availability when main main steam tunnel, failure of main turbine trip due to main steam steam tunnel high steam cooling will not result in a tunnel (MST) high temperature temperature exists. MST temperature of 185°F or when PSW is unavailable and no greater. Therefore, it will not result steam line break has occurred. in an initiation of the MST high temperature isolation logic.

229 Increase the training emphasis Increased availability of Retain In accordance with GDC 13, No and provide additional control the SSW pump house (Phase II damper status is indicated in the room indication on the operational ventilation system. SAMA 58) main control room. In addition, there status of the SSW pump house is a high temperature alarm in the ventilation system. This will allow main control room.

operators to manually open the Alarm 04-1-02-1H13-P870 provides pump house dampers, which can an alarm, but the actions could be provide adequate ventilation such expanded to accomplish a more that pump failures would not occur. robust mitigation of this condition.

230 Increase operator training for Increased time available Retain No specific operator training is in No alternate operation of the low for recovery actions for low (Phase II place to address this condition.

pressure ECCS pumps (LPCI and pressure ECCS when a SAMA 59)

LPCS) for loss of SSW scenarios. loss of SSW occurs.

E.2-21

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-1 (Continued)

Phase I SAMAs Related to IPE and IPEEE Insights Phase I Credited SAMA Result of Potential Screening SAMA Title SAMA Disposition in PSA ID Enhancement Results Model Number 231 Revise the containment flooding Limit one of the major #3 - Already GGNS contributed this IPE insight Yes portion of the Emergency contributors to the source installed to the BWR Owners Group Severe Procedures to remove or modify term released. Accident Subcommittee. GGNS has the step requiring MSIV venting. already implemented the current SAGs on RPV venting.

232 Install a backup power supply to Hydrogen igniter #3 - Already GGNS has two hydrogen No the hydrogen igniters. operability during station Installed recombiners, each powered from a blackout. 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.

233 Install an additional method of Increased decay heat Retain GGNS utilizes the containment No removing heat from the removal capability (Phase II spray and RHR suppression pool containment. SAMA 60) 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.

E.2-22

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-1 (Continued)

Phase I SAMAs Related to IPE and IPEEE Insights Phase I Credited SAMA Result of Potential Screening SAMA Title SAMA Disposition in PSA ID Enhancement Results Model Number 234 Install a backup water supply and Alternate water supply for Retain GGNS has a high pressure core No pumping capability that is containment spray/vessel (Phase II spray system, which is powered independent of normal and injection SAMA 61) from an independent (Division 3) emergency AC power. power supply; however, a backup supply will be investigated per the IPE recommendations.

235 Extend the battery depletion time Enhanced reactor #2 - Similar ADS and Non-ADS relief valves are No for the relief valves. pressure vessel item is all dependent on DC power and depressurization system addressed instrument air. Extended DC power reliability under other to the relief valves will allow longer proposed operation during a loss of DC SAMAs battery chargers.

Similar to Phase II SAMAs 1, 3, and 27.

236 Implement the latest revision of the Improved likelihood of #3 - Already GGNS currently utilizes Revision 2 Yes BWR Owners Group emergency success of operator Installed of the BWROG EPGs.

procedure guidelines (EPGs). actions taken in response to abnormal conditions.

237 Increase maintenance on drainage Prevent deterioration of #3 - Already GGNS has increased the No structures. Maintenance should site conditions. installed maintenance on drainage include cleaning of culverts, structures.

concrete repair, and removal of vegetation/debris which could obstruct flow.

E.2-23

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-1 (Continued)

Phase I SAMAs Related to IPE and IPEEE Insights Phase I Credited SAMA Result of Potential Screening SAMA Title SAMA Disposition in PSA ID Enhancement Results Model Number 238 Plant procedures currently require Reduce leakage from #3 - Already GGNS has revised the plant flood No plant staff to insure that plant doors flooding through an open installed mitigation procedure.

are closed during severe weather door.

and in the event of plant flooding (Implicitly including former Unit 2 doors). Revise procedures to explicitly include at-grade former Unit 2 doors.

239 Revise procedures to periodically Reduce the consequences #3 - Already GGNS has created an inspection No inspect roof drains and overflows of a flood. installed procedure for roof drains, roof to ensure they are not blocked. drainage system, and roof overflows.

E.2-24

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-1 (Continued)

Phase I SAMAs Related to IPE and IPEEE Insights Phase I Credited SAMA Result of Potential Screening SAMA Title SAMA Disposition in PSA ID Enhancement Results Model Number 240 Remove the wooden foot bridge Improve site drainage/ #1 - N/A The IPEEE showed the risk from No crossing the northwest ditch near external flood protection. external flooding at GGNS is minor.

its upstream end. 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.

E.2-25

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-1 (Continued)

Phase I SAMAs Related to IPE and IPEEE Insights Phase I Credited SAMA Result of Potential Screening SAMA Title SAMA Disposition in PSA ID Enhancement Results Model Number 241 Remove the 15" corrugated metal Improve site drainage/ #1 - N/A The IPEEE showed the risk from No pipe located in the small auxiliary external flood protection. external flooding at GGNS is minor.

ditch parallel to the northwest ditch Thus this potential modification is (at the same approximate location assumed not to be cost beneficial, as the duct bank crossing the which follows the same assumption northwest ditch). Re-grade the in the NRC safety evaluation report.

area to provide a gradual transition In May 2011, NRC Inspectors between the yard upstream and verified that the plant grade is 132.5 the auxiliary ditch. 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.

E.2-26

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-1 (Continued)

Phase I SAMAs Related to IPE and IPEEE Insights Phase I Credited SAMA Result of Potential Screening SAMA Title SAMA Disposition in PSA ID Enhancement Results Model Number 242 Re-hang the security fence gates Improve site drainage/ #1 - N/A The IPEEE showed the risk from No west of the control building to external flood protection. external flooding at GGNS is minor.

insure that approximately 5" of gap Thus this potential modification is exists between the gate and the assumed not to be cost beneficial, road. 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.

E.2-27

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-1 (Continued)

Phase I SAMAs Related to IPE and IPEEE Insights Phase I Credited SAMA Result of Potential Screening SAMA Title SAMA Disposition in PSA ID Enhancement Results Model Number 243 Grade down and remove the Improve site drainage/ #1 - N/A The IPEEE showed the risk from No access road, the raised berm external flood protection. external flooding at GGNS is minor.

parallel to the access road, and Thus this potential modification is curbs adjacent to the access road assumed not to be cost beneficial, as necessary where they cross which follows the same assumption Culvert No.1, such that elevations in the NRC safety evaluation report.

above the culvert do not exceed In May 2011, NRC Inspectors 132.7 ft. MSL. 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.

E.2-28

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-1 (Continued)

Phase I SAMAs Related to IPE and IPEEE Insights Phase I Credited SAMA Result of Potential Screening SAMA Title SAMA Disposition in PSA ID Enhancement Results Model Number 244 Replace the C8x11.5 channel Improve site drainage/ #1 - N/A The IPEEE showed the risk from No forming the flood barrier across the external flood protection. external flooding at GGNS is minor.

SSW A equipment hatch opening Thus this potential modification is with another member having a assumed not to be cost beneficial, minimum depth of approximately which follows the same assumption 13". 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.

245 Modify the piping systems to Reduce vulnerability to a #3 - Already The grout was removed and the No account for the grouted condition seismic event. installed pipe support at the penetration was for the penetration of the standby modified to coincide with the design service water (SSW) piping in the basis piping analysis assumption.

control building.

E.2-29

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

1. DC Power Eliminates all SBO 13.6% 16.5% 13.6% $115,656 $346,968 cutsets 1 - Provide additional CNS estimate. $500,000 Not cost DC battery capacity effective 2 - Replace lead-acid CNS estimate. $1,000,000 Not cost batteries with fuel effective cells 11 - Portable CNS had different $714,000 Not cost generator for direct cost estimates for effective current (DC) power: the portable This SAMA involves generator to supply the use of a portable the charger and the generator to supply portable generator to DC power to the supply a panel battery chargers because they had an during a station existing generator blackout. big enough to supply the charger, but not big enough to supply a panel.

(cont. below)

E.2-30

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion (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 CNS estimate. $714,000 Not cost generator for direct effective current (DC) power:

This SAMA involves the use of a portable generator to supply DC power to the individual panels during a station blackout.

E.2-31

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion 15 - Use DC GGNS SAMA 11 and $1,428,000 Not cost generators to provide SAMA 12 estimate effective power to operate the that one generator switchyard power would cost control breakers while ~$714,000. This a 480-V AC generator SAMA recommends could supply the air addition of at least compressors for two generators.

breaker support. Thus, GGNS SAMA 15 cost estimate should be at least double that for SAMA 11 or 12.

2. Improve Charger Failure of chargers 1.4% 2.2% 2.3% $13,598 $40,793 Reliability contribution to zero.

3 - Add battery CNS estimate. $90,000 Not cost charger to existing effective DC system 13 - Proceduralize CNS estimate. $50,000 Not cost battery charger high- effective voltage shutdown circuit inhibit E.2-32

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

3. Add DC System Eliminate failure of 7.6% 11.6% 11.8% $73,056 $219,169 Cross-ties DC power gates.

4 - Provide DC bus CNS estimate. $300,000 Not cost cross-ties effective

4. Increase Eliminated failure of 17.5% 21.2% 18.5% $149,396 $448,189 Availability of On- DG11, DG12, and Site AC Power DG13 to their AC Busses 5 - Provide an CNS estimate. $20,000,000 Not cost additional diesel effective generator 8 - Install a gas CNS estimate. $2,000,000 Not cost turbine generator with effective tornado protection

5. Improve AC Eliminated the loss 20.4% 25.6% 23.2% $177,524 $532,571 Power of the 4.16-kV buses 6 - Improve 4.16-kV CNS estimate. $656,000 Not cost bus cross-tie ability effective E.2-33

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion 17 - Provide alternate Modification of the $656,000 Not cost feeds to essential AC system to allow effective loads directly from an alignment of alternate emergency alternate feeds to the bus 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.

6. Reduce Loss of Eliminate the 3.1% 3.7% 3.1% $26,087 $78,261 Off-Site Power weather centered During Severe loss of off-site power Weather initiating event.

7 - Install an CNS estimate. $2,485,000 Not cost additional, buried off- effective site power source.

E.2-34

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

7. Provide Backup Eliminated failure of 1.9% 2.5% 1.9% $16,515 $49,545 EDG Cooling SW cooling to the EDGs 9 - Use fire water Hardware $100,000 Not cost system as backup modification range effective source for diesel estimate.

cooling 10 - Add new backup CNS estimate. $2,000,000 Not cost source of diesel effective cooling

8. Increase EDG Eliminated failure of 3.3% 4.6% 4.5% $30,348 $91,044 Reliability EDGs to run 14 - Provide a CNS estimate. $100,000 Not cost portable EDG fuel oil effective transfer pump E.2-35

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

9. Improve DG Eliminated the 0.3% 0.3% 0.2% $2,181 $6,542 Reliability common cause failure (CCF) contribution of failure to start EDGs 16 - Provide a Hardware $100,000 Not cost diverse swing diesel modification range effective generator air start estimate.

compressor

10. Reduce Plant- Removed the 9.1% 10.7% 8.9% $76,556 $229,668 Centered Loss of contribution of plant-Off-Site Power and switchyard-centered events 18 - Protect CNS estimate. $780,000 Not cost transformers from effective failure E.2-36

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

11. Redundant Eliminated failure of 1.1% 1.8% 1.8% $10,766 $32,297 Power to Torus Hard power to Pipe Vent (THPV) containment vents Valves 19 - Provide CNS estimate. $714,000 Not cost redundant power to effective direct torus hard pipe vent valves to improve the reliability of the direct torus vent valves and enhance the containment heat removal capability.

E.2-37

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

12. High Pressure Eliminated failure of 77.8% 61.8% 60.2% $594,912 $1,784,736 Injection System the HPCS 20 - Install an Recent BWR cost $8,800,000 Not cost independent active or estimates for this effective passive high pressure SAMA are ~$2M at injection system 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.

E.2-38

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion 61 - Install a backup Plant-specific cost $6,409,949 Not cost water supply and estimate. effective pumping capability that is independent of normal and emergency AC power

13. Extend RCIC Eliminated failure of 1.0% 1.6% 1.7% $10,031 $30,093 Operation trip due to pressure 21 - Raise HPCI/ CNS estimate. $200,000 Not cost RCIC backpressure effective trip set points [HPCI backpressure trip setpoint has already been raised. This SAMA will evaluate raising the RCIC backpressure trip set point].

E.2-39

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

14. Improve ADS Eliminated failure of 45.9% 16.3% 16.0% $299,106 $897,317 System ADS valves 22 - Modify automatic Plant-specific cost $1,176,850 Not cost depressurization estimate. effective system components to improve reliability

[This SAMA will add larger accumulators thus increasing reliability during SBOs].

15. Improve ADS Eliminated failure of 20.8% 5.3% 4.8% $129,383 $388,150 Signals the SRV failing to open 23 - Add signals to CNS estimate. $1,500,000 Not cost open safety relief effective valves automatically in an MSIV closure transient.

E.2-40

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

16. Low Pressure Eliminated failure of 22.9% 39.5% 38.3% $229,965 $689,896 Injection System the LPCI and LPCS 24 - Add a diverse CNS estimate. $8,800,000 Not cost low pressure injection effective system.

17. ECCS Low Eliminated ECCS 1.0% 1.6% 1.7% $10,031 $30,093 Pressure Interlock permissives and interlock failure 25 - Install a bypass CNS estimate. $1,000,000 Not cost switch to allow effective 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.

E.2-41

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

18. RHR Heat Eliminated failure of 18.5% 37.4% 39.9% $205,223 $615,669 Exchangers SSW to provide cooling to the RHR heat exchangers 26 - Implement Pilgrim estimate. $1,950,000 Not cost modifications to allow effective manual alignment of the fire water system to RHR heat exchangers.

19. Emergency Eliminated failure of 3.6% 6.7% 7.0% $37,903 $113,708 Service Water service water pumps System Reliability 27 - Add a service CNS estimate. $5,900,000 Not cost water pump to effective increase availability of cooling water

20. Main Feedwater Eliminated failure to 19.3% 20.5% 20.6% $162,050 $486,149 System Reliability inject from feedwater 28 - Add a motor- CNS estimate. $1,650,000 Not cost driven feed water effective pump E.2-42

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

21. Increase Eliminated failure of 22.9% 17.8% 18.2% $175,400 $526,200 Availability of Room room cooling to Cooling LPCS, HPCS, SSW and safeguard switchgear battery rooms 29 - Provide a CNS estimate. $2,202,725 Not cost redundant train or effective means of ventilation

22. Increase Eliminated failure of 9.2% 10.6% 8.5% $75,988 $227,963 Availability of the diesel generator DG System Through rooms HVAC HVAC Improvements 30 - Add a diesel CNS estimate. $1,304,700 Not cost building high effective temperature alarm or redundant louver and thermostat.

32 - Diverse EDG Cost for Phase II $300,000 Not cost HVAC logic SAMAs 4 and 31 is effective used because the modifications are similar in scope.

E.2-43

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion 33 - Install additional CNS estimate. $6,000,000 Not cost fan and louver pair for effective EDG heating, ventilation, and air conditioning

23. Increased Eliminated failure of 1.0% 1.6% 1.7% $10,031 $30,093 reliability of HPCI power to the HPCS and RCIC room pump room cooler.

cooling (RCIC pump continued operation is not dependent on room cooling.)

31 - Create ability to CNS estimate. $300,000 Not cost switch HPCI and Similar to SAMA 4, effective RCIC room fan power provide DC bus supply to DC in an cross-ties.

SBO event.

E.2-44

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

24. Increase Eliminated failure of 14.9% 20.2% 21.3% $137,842 $413,527 Reliability of the instrument air Instrument Air 34 - Modify CNS estimate. More $1,200,000 Not cost procedure/hardware than just procedure. effective to provide ability to align diesel power to more air compressors 35 - Replace service CNS estimate. $1,394,598 Not cost and instrument air effective compressors with more reliable compressors which have self-contained air cooling by shaft-driven fans

25. Backup Nitrogen Eliminated operator 5.5% 3.7% 3.8% $40,614 $121,841 to SRV failure to install air bottles 36 - Install nitrogen Plant-specific cost $1,722,706 Not cost bottles as backup gas estimate. effective supply for safety relief valves.

E.2-45

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

26. Improve Eliminated failure of 46.1% 16.4% 16.1% $300,631 $901,893 Availability of SRVs non-ADS SRVs and MSIVs 37 - Improve SRV CNS estimate. $1,500,000 Not cost and MSIV pneumatic effective components.

27. Improve Eliminated the failure 18.5% 37.4% 39.9% $205,223 $615,669 Suppression Pool of flow to the RHR Cooling heat exchangers 38 - Install an CNS estimate. $5,800,000 Not cost independent method effective of suppression pool cooling.

28. Increase Eliminated failure of 26.6% 51.6% 54.7% $288,437 $865,312 Availability of cooled flow from Containment Heat RHR pump A and B Removal 39 - Procedural Procedural range $25,000 Retain change to cross-tie estimate.

open cycle cooling system to enhance containment spray system E.2-46

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion 41 - Use the fire Similar to Phase II $1,950,000 Not cost water system as a SAMA 26, implement effective backup source for the modifications to drywell spray system allow manual alignment of the fire water system to RHR heat exchangers.

29. Decay Heat Eliminated failure of 26.6% 51.6% 54.7% $288,550 $865,649 Removal Capability RHR spray

- Drywell Spray 40 - Install a passive CNS estimate. $5,800,000 Not cost drywell spray system effective to provide redundant drywell spray method.

30. Increase Eliminated failure of 11.3% 16.8% 17.4% $107,899 $323,696 Availability of the HPCS and RCIC CST suction 42 - Enhance Procedure with $200,000 Retain procedures to refill engineering and CST from training range demineralized water estimate.

or service water system.

E.2-47

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

31. Filtered Vent to Reduced the 0.0% 26.4% 34.3% $80,920 $242,759 Increase Heat baseline accident Removal Capacity progression source for Non-ATWS terms by a factor of 2 Events 43 - Install a filtered CNS estimate. $1,500,000 Not cost containment vent to effective provide fission product scrubbing

32. Reduce Eliminated failure of 15.9% 20.7% 20.2% $142,455 $427,365 Hydrogen Ignition hydrogen igniters 44 - Provide post- Plant-specific cost $2,665,123 Not cost accident containment estimate. effective inerting capability.

45 - Install a passive Monticello (SAMA $760,000 Not cost hydrogen control 10) estimated that effective system. this modification would cost

~$760,000.

E.2-48

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

33. Controlled Eliminated failure of 2.9% 5.4% 5.8% $31,080 $93,240 Containment air-operated valves Venting to open 46 - Provide passive CNS estimate. $1,000,000 Not cost overpressure relief by effective changing the containment vent valves to fail open and improving the strength of the rupture disk 47 - Enable manual Oyster Creek (SAMA $150,000 Not cost operation of all 84) estimated that it effective containment vent would cost $150,000 valves via local to add handwheels in controls the reactor building to open AOVs in the current vent path.

34. ISLOCA Removed all < 0.1% < 0.1% < 0.1% $77 $231 ISLOCA initiators 48 - Increase CNS estimate. $100,000 Not cost frequency of valve effective leak testing to reduce ISLOCA frequency E.2-49

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion 50 - Revise EOPs to CNS estimate. $50,000 Not cost improve ISLOCA effective identification 51 - Improve CNS estimate. $112,000 Not cost operator training on effective ISLOCA coping

35. MSIV Design Eliminated failure of 1.0% 1.6% 1.7% $10,031 $30,093 the MSIVs to close or remain closed 49 - Improve MSIV CNS estimate. $1,000,000 Not cost design to decrease effective the likelihood of containment bypass scenarios.

E.2-50

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

36. SLC System Eliminated failure to 1.1% 1.7% 1.7% $10,616 $31,849 initiate SLC and failures of alternate boron injection (ABI) 52 - Increase boron CNS estimate. $50,000 Not cost concentration in the effective SLC system

[Reduced time required to achieve shutdown provides increased margin in the accident timeline for successful initiation of SLC]

E.2-51

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

37. SRV Reseat Eliminated the 3.1% 4.3% 4.5% $29,108 $87,324 initiator for SRVs inadvertently being open and basic events for stuck open SRVs 53 - Increase safety CNS estimate. $2,200,000 Not cost relief valve (SRV) effective 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

38. Add Fire Eliminated fire CDF n/a n/a n/a $34,115 $102,345 Suppression1 from the critical switchgear rooms.

54 - Add automatic CNS estimate. $375,000 Not cost fire suppression effective systems to the dominant fire zones E.2-52

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

39. Reduce Risk Eliminate fire CDF n/a n/a n/a $140,174 $420,521 from Fires that from the main control Require Control room.

Room Evacuation(1) 55 - Upgrade the CNS estimate. $786,991 Not cost ASDS panel to effective include additional system controls for opposite division.

40. Large Break Eliminated Large 7.1% 16.5% 17.5% $316,124 $948,372 LOCA Break LOCA 56 - Provide digital Duane Arnold $2,000,000 Not cost large break LOCA estimated that this effective protection to identify modification would symptoms/precursors cost at least $2M.

of a large break LOCA (a leak before break)

E.2-53

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

41. Trip/Shutdown Reducing all initiating 8.0% 6.7% 6.9% $62,372 $187,117 Risk events except pipe breaks, floods, and LOSP by a factor of 2

57 - Generation Risk CNS estimate. $500,000 Not cost Assessment effective implementation into plant activities (trip/

shutdown risk modeling).

42. Increase Eliminated failure of 1.6% 2.2% 2.3% $15,071 $45,212 Availability of SSW SSW Pump House Pump House Ventilation Ventilation System 58 - Increase the Hardware $100,000 Not cost training emphasis and modification range effective provide additional estimate.

control room indication on the operational status of SSW pump house ventilation system.

E.2-54

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

43. Increase Eliminated failure of 4.1% 6.5% 6.8% $40,452 $121,357 recovery time of SSW to the LPCS ECCS upon loss of room cooler SSW 59 - Increase Procedure with $50,000 Retain operator training for training range alternating operation estimate.

of the low pressure ECCS pumps (LPCI and LPCS) for loss of SSW scenarios.

44. Additional Eliminated failure of 27.5% 53.2% 56.3% $298,121 $894,362 Containment Heat suppression pool Removal cooling and containment spray systems 60 - Install an Plant-specific cost $4,352,023 Not cost additional method of estimate. effective heat removal from containment.

E.2-55

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-2 (Continued)

Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Analysis Case Internal Internal and (bold) and External SAMA Number and CDF PDR OECR External Benefit with GGNS Cost Title Assumptions Reduction Reduction Reduction Benefit Uncertainty Estimate Conclusion

45. Improve RHR Eliminated failure of 3.6% 7.8% 8.3% $41,340 $124,019 Heat Exchanger RHR HX Cooler inlet Availability and outlet valves 62 - Add a bypass Plant-specific cost $2,831,652 Not cost around the RHR HX estimate. effective inlet and outlet valves

46. Improve RCIC Eliminated the failure 4.7% 1.9% 1.6% $30,894 $92,683 Lube Oil Cooling to cool RCIC lube oil 63 - Add a redundant Hardware $100,000 Not cost RCIC lube oil cooling modification range effective path. estimate.

1. These analysis cases only impact external events and have been evaluated differently as shown in Section E.2.3.

E.2-56

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-3 Sensitivity Analysis Results Sensitivity Case 1, Sensitivity Case 2, Internal and Internal and Internal and External Benefit, External Benefit, External Benefit, Analysis Case (bold) 20 yrs Remaining, 33 yrs Remaining, 20 yrs Remaining, GGNS Cost SAMA Number and Title 7% Discount Rate 7% Discount Rate 3% Discount Rate Estimate

1. DC Power $115,656 $171,775 $144,423 1 - Provide additional DC battery capacity $500,000 2 - Replace lead-acid batteries with fuel cells $1,000,000 11 - Portable generator for direct current (DC) power: $714,000 This SAMA involves the use of a portable generator to supply DC power to the battery chargers during a station blackout.

12 - Portable generator for direct current (DC) power: $714,000 This SAMA involves the use of a portable generator to supply DC power to the individual panels during a station blackout.

15 - Use DC generators to provide power to operate $1,428,000 the switchyard power control breakers while a 480-V AC generator could supply the air compressors for breaker support.

2. Improve Charger Reliability $13,598 $19,619 $17,276 3 - Add battery charger to existing DC system $90,000 13 - Proceduralize battery charger high-voltage $50,000 shutdown circuit inhibit

3. Add DC System Cross-Ties $73,056 $105,875 $92,577 4 - Provide DC bus cross-ties $300,000 E.2-57

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-3 (Continued)

Sensitivity Analysis Results Sensitivity Case 1, Sensitivity Case 2, Internal and Internal and Internal and External Benefit, External Benefit, External Benefit, Analysis Case (bold) 20 yrs Remaining, 33 yrs Remaining, 20 yrs Remaining, GGNS Cost SAMA Number and Title 7% Discount Rate 7% Discount Rate 3% Discount Rate Estimate

4. Increase Availability of On-Site AC Power $149,396 $221,380 $186,814 5 - Provide an additional diesel generator $20,000,000 8 - Install a gas turbine generator with tornado $2,000,000 protection

5. Improve AC Power $177,524 $262,069 $222,495 6 - Improve 4.16-kV bus cross-tie ability $656,000 17 - Provide alternate feeds to essential loads directly $656,000 from an alternate emergency bus

6. Reduce Loss of Off-Site Power During Severe $26,087 $38,786 $32,554 Weather 7 - Install an additional, buried off-site power source. $2,485,000

7. Provide Backup EDG Cooling $16,515 $24,490 $20,642 9 - Use fire water system as backup source for diesel $100,000 cooling 10 - Add new backup source of diesel cooling $2,000,000

8. Increase EDG Reliability $30,348 $44,328 $38,279 14 - Provide a portable EDG fuel oil transfer pump $100,000 E.2-58

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-3 (Continued)

Sensitivity Analysis Results Sensitivity Case 1, Sensitivity Case 2, Internal and Internal and Internal and External Benefit, External Benefit, External Benefit, Analysis Case (bold) 20 yrs Remaining, 33 yrs Remaining, 20 yrs Remaining, GGNS Cost SAMA Number and Title 7% Discount Rate 7% Discount Rate 3% Discount Rate Estimate

9. Improve DG reliability $2,181 $3,249 $2,718 16 - Provide a diverse swing diesel generator air start $100,000 compressor

10. Reduce Plant-Centered Loss of Off-Site Power $76,556 $113,849 $95,522 18 - Protect transformers from failure $780,000

11. Redundant Power to Torus Hard Pipe Vent $10,766 $15,502 $13,694 (THPV) Valves 19 - Provide redundant power to direct torus hard pipe $714,000 vent valves to improve the reliability of the direct torus vent valves and enhance the containment heat removal capability.

12. High Pressure Injection System $594,912 $901,576 $733,645 20 - Install an independent active or passive high $8,800,000 pressure injection system 61 - Install a backup water supply and pumping $6,409,949 capability that is independent of normal and emergency AC power E.2-59

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-3 (Continued)

Sensitivity Analysis Results Sensitivity Case 1, Sensitivity Case 2, Internal and Internal and Internal and External Benefit, External Benefit, External Benefit, Analysis Case (bold) 20 yrs Remaining, 33 yrs Remaining, 20 yrs Remaining, GGNS Cost SAMA Number and Title 7% Discount Rate 7% Discount Rate 3% Discount Rate Estimate

13. Extend RCIC Operation $10,031 $14,448 $12,757 21 - Raise HPCI/RCIC backpressure trip set points $200,000

[HPCI backpressure trip setpoint has already been raised. This SAMA will evaluate raising the RCIC backpressure trip set point].

14. Improve ADS System $299,106 $469,925 $360,320 22 - Modify automatic depressurization system $1,176,850 components to improve reliability [This SAMA will add larger accumulators thus increasing reliability during SBOs].

15. Improve ADS Signals $129,383 $205,503 $154,719 23 - Add signals to open safety relief valves $1,500,000 automatically in an MSIV closure transient.

16. Low Pressure Injection System $229,965 $331,005 $292,574 24 - Add a diverse low pressure injection system. $8,800,000

17. ECCS Low Pressure Interlock $10,031 $14,448 $12,757 25 - Install a bypass switch to allow operators to $1,000,000 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.

E.2-60

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-3 (Continued)

Sensitivity Analysis Results Sensitivity Case 1, Sensitivity Case 2, Internal and Internal and Internal and External Benefit, External Benefit, External Benefit, Analysis Case (bold) 20 yrs Remaining, 33 yrs Remaining, 20 yrs Remaining, GGNS Cost SAMA Number and Title 7% Discount Rate 7% Discount Rate 3% Discount Rate Estimate

18. RHR Heat Exchangers $205,223 $290,595 $263,557 26 - Implement modifications to allow manual $1,950,000 alignment of the fire water system to RHR heat exchangers.

19. Emergency Service Water System Reliability $37,903 $54,031 $48,491 27 - Add a service water pump to increase availability $5,900,000 of cooling water

20. Main Feedwater System Reliability $162,050 $241,055 $202,163 28 - Add a motor-driven feed water pump $1,650,000

21. Increase Availability of Room Cooling $175,400 $265,739 $216,342 29 - Provide a redundant train or means of ventilation $2,202,725

22. Increase Availability of the DG System through $75,988 $113,283 $94,669 HVAC Improvements 30 - Add a diesel building high temperature alarm or $1,304,700 redundant louver and thermostat.

32 - Diverse EDG HVAC logic $300,000 33 - Install additional fan and louver pair for EDG $6,000,000 heating, ventilation, and air conditioning E.2-61

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-3 (Continued)

Sensitivity Analysis Results Sensitivity Case 1, Sensitivity Case 2, Internal and Internal and Internal and External Benefit, External Benefit, External Benefit, Analysis Case (bold) 20 yrs Remaining, 33 yrs Remaining, 20 yrs Remaining, GGNS Cost SAMA Number and Title 7% Discount Rate 7% Discount Rate 3% Discount Rate Estimate

23. Increased Reliability of HPCI and RCIC Room $10,031 $14,448 $12,757 Cooling 31 - Create ability to switch HPCI and RCIC room fan $300,000 power supply to DC in an SBO event.

24. Increase Reliability of Instrument Air $137,842 $201,172 $173,951 34 - Modify procedure/hardware to provide ability to $1,200,000 align diesel power to more air compressors 35 - Replace service and instrument air compressors $1,394,598 with more reliable compressors which have self-contained air cooling by shaft-driven fans

26. Backup Nitrogen to SRV $40,614 $62,050 $49,828 36 - Install nitrogen bottles as backup gas supply for $1,722,706 safety relief valves.

26. Improve Availability of SRVs and MSIVs $300,631 $472,257 $362,190 37 - Improve SRV and MSIV pneumatic components. $1,500,000

27. Improve Suppression Pool Cooling $205,223 $290,595 $263,557 38 - Install an independent method of suppression pool $5,800,000 cooling.

E.2-62

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-3 (Continued)

Sensitivity Analysis Results Sensitivity Case 1, Sensitivity Case 2, Internal and Internal and Internal and External Benefit, External Benefit, External Benefit, Analysis Case (bold) 20 yrs Remaining, 33 yrs Remaining, 20 yrs Remaining, GGNS Cost SAMA Number and Title 7% Discount Rate 7% Discount Rate 3% Discount Rate Estimate

28. Increase Availability of Containment Heat $288,437 $409,794 $369,722 Removal 39 - Procedural change to cross-tie open cycle cooling $25,000 system to enhance containment spray system 41 - Use the fire water system as a backup source for $1,950,000 the drywell spray system

29. Decay Heat Removal Capability - Drywell Spray $288,550 $409,953 $369,866 40 - Install a passive drywell spray system to provide $5,800,000 redundant drywell spray method.

30. Increase Availability of the CST $107,899 $156,536 $136,643 42 - Enhance procedures to refill CST from $200,000 demineralized water or service water system.

31. Filtered Vent to Increase Heat Removal Capacity $80,920 $96,745 $113,074 for Non-ATWS Events 43 - Install a filtered containment vent to provide fission $1,500,000 product scrubbing

32. Reduce Hydrogen Ignition $142,455 $209,206 $179,104 44 - Provide post-accident containment inerting $2,665,123 capability.

45 - Install a passive hydrogen control system. $760,000 E.2-63

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-3 (Continued)

Sensitivity Analysis Results Sensitivity Case 1, Sensitivity Case 2, Internal and Internal and Internal and External Benefit, External Benefit, External Benefit, Analysis Case (bold) 20 yrs Remaining, 33 yrs Remaining, 20 yrs Remaining, GGNS Cost SAMA Number and Title 7% Discount Rate 7% Discount Rate 3% Discount Rate Estimate

33. Controlled Containment Venting $31,080 $44,274 $39,779 46 - Provide passive overpressure relief by changing $1,000,000 the containment vent valves to fail open and improving the strength of the rupture disk 47 - Enable manual operation of all containment vent $150,000 valves via local controls

34. ISLOCA $77 $118 $95 48 - Increase frequency of valve leak testing to reduce $100,000 ISLOCA frequency 50 - Revise EOPs to improve ISLOCA identification $50,000 51 - Improve operator training on ISLOCA coping $112,000

35. MSIV Design $10,031 $14,448 $12,757 49 - Improve MSIV design to decrease the likelihood of $1,000,000 containment bypass scenarios.

36. SLC System $10,616 $15,376 $13,458 52 - Increase boron concentration in the SLC system $50,000

[Reduced time required to achieve shutdown provides increased margin in the accident timeline for successful initiation of SLC]

E.2-64

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-3 (Continued)

Sensitivity Analysis Results Sensitivity Case 1, Sensitivity Case 2, Internal and Internal and Internal and External Benefit, External Benefit, External Benefit, Analysis Case (bold) 20 yrs Remaining, 33 yrs Remaining, 20 yrs Remaining, GGNS Cost SAMA Number and Title 7% Discount Rate 7% Discount Rate 3% Discount Rate Estimate

37. SRV Reseat $29,108 $42,416 $36,767 53 - Increase safety relief valve (SRV) reseat reliability $2,200,000 to address the risk associated with dilution of boron caused by the failure of the SRVs to reseat after standby liquid control (SLC) injection

38. Add Fire Suppression1 N/A N/A N/A 54 - Add automatic fire suppression systems to the $375,000 dominant fire zones

39. Reduce Risk from Fires that Require Control N/A N/A N/A Room Evacuation1 55 - Upgrade the ASDS panel to include additional $786,991 system controls for opposite division.

40. Large Break LOCA $316,124 $463,652 $380,827 56 - Provide digital large break LOCA protection to $2,000,000 identify symptoms/precursors of a large break LOCA (a leak before break)

41. Trip/Shutdown Risk $62,372 $94,032 $77,170 57 - Generation Risk Assessment implementation into $500,000 plant activities (trip/shutdown risk modeling).

E.2-65

Grand Gulf Nuclear Station Applicants Environmental Report Operating License Renewal Stage Table E.2-3 (Continued)

Sensitivity Analysis Results Sensitivity Case 1, Sensitivity Case 2, Internal and Internal and Internal and External Benefit, External Benefit, External Benefit, Analysis Case (bold) 20 yrs Remaining, 33 yrs Remaining, 20 yrs Remaining, GGNS Cost SAMA Number and Title 7% Discount Rate 7% Discount Rate 3% Discount Rate Estimate

42. Increase Availability of SSW Pump House $15,071 $21,998 $19,017 Ventilation System 58 - Increase the training emphasis and provide $100,000 additional control room indication on the operational status of SSW pump house ventilation system.

43. Increase Recovery Time of ECCS upon Loss of $40,452 $58,438 $51,357 SSW 59 - Increase operator training for alternating operation $50,000 of the low pressure ECCS pumps (LPCI and LPCS) for loss of SSW scenarios.

44. Additional Containment Heat Removal $298,121 $423,739 $382,038 60 - Install an additional method of heat removal from $4,352,023 containment.

45. Improve RHR Heat Exchanger Availability $41,340 $58,200 $53,263 62 - Add a bypass around the RHR HX inlet and outlet $2,831,652 valves

46. Improve RCIC Lube Oil Cooling $30,894 $48,447 $37,264 63 - Add a redundant RCIC lube oil cooling path. $100,000

1. These analysis cases only impact external events and have been evaluated differently as shown in Section E.2.3.

E.2-66

GNRO-2011/00093, Applicants Environmental Report, Operating License Renewal Stage, Attachment E - Severe Accident Mitigation Alternatives Analysis

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GNRO-2011/00093, Applicants Environmental Report, Operating License Renewal Stage, Attachment E - Severe Accident Mitigation Alternatives Analysis

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