ML110600910: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(Created page by program invented by StriderTol)
Line 17: Line 17:


=Text=
=Text=
{{#Wiki_filter:Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating 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 .....................................
{{#Wiki_filter:Nuclear Regulatory Commission Exhibit # - NRC000001-00-BD01 Docket # - 05000293 Identified: 02/22/2011 Admitted: 02/22/2011        Withdrawn:
E.1-1 E.1.2 PSA Model -Level 2 Analysis .....................................
Rejected:                  Stricken:
E.1-27 E.1.2.1 Containment Performance Analysis .............................
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Attachment E Severe Accident Mitigation Alternatives Analysis Attachment E contains the following sections.
E.1-27 E.1.2.2 Radionuclide Analysis ........................
E.1 - Evaluation of PSA Model E.2 - Evaluation of SAMA Candidates
E.1-33 E. 1.2.2.1 Introduction
 
....................................
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                                      Pilgrim LR Proceeding SAMA Analysis                                                               50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's 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.1-27 E.1.2.1 Containment Performance Analysis .............................                                                        E.1-27 E.1.2.2 Radionuclide Analysis                          ........................                                             E.1-33 E. 1.2.2.1 Introduction  ....................................                                                           E. 1-33 E.1.2.2.2 Timing of Release         ........................                                                            E.1-33 E.1.2.2.3  Magnitude of Release ......................                                                                   E.1-34 E.1.2.2.4  Release Category Bin Assignments .......                                    ................. E.1-34 E.1.2.2.5  Mapping of Level 1 Results into the Various Release Categories . E.1-35 E.1.2.2.6  Collapsed Accident Progression Bins Source Terms ....                                               ....... E.1-43 E.1.2.2.7 Release Magnitude Calculations ..........................                                                     E.1-52 E.1.3 IPEEE Analysis                       ..........................                                                           E.1-52 E.1.3.1 Seismic Analysis ..........................                                                                           E.1-52 E.1.3.2 Fire Analysis .......................                                       .         .                 .         . E.1-52 E.1.3.3 Other External Hazards ......................                                                                         E.1-54 E.1.4 PSA Model Peer Review and Difference between Current PSA Model and 1995 Update IPE                      ..                                   .....................                         E.1-54 E.1.4.1 PSA Model Peer Review ........................                                           ....                       E.1 -54 E.1.4.2 Major Differences between the Updated IPE PSA Model and 1995 Update IPE Model .................                                                                             E.1-55 E.1.4.2.1 Core Damage - Comparison to the PNPS 1995 Update IPE Model         .........................                                                         E.1-55 E.1.4.2.2 Containment Performance                    -  Comparison to the Original PNPS IPE Model ..........                                                                                   E.1-59 E.1.5 The MACCS2 Model - Level 3 Analysis                      .          .          ............................               E.1-60 E.1.5.1 Introduction                    .................                                               I........           E.1-60 E.1.5.2 Input                  ................                                                                             E.1-60 E.1.5.2.1 Projected Total Population by Spatial Element .....                                          .......... E.1-61 E.1.5.2.2 Land Fraction ....................                                   ........                                 E.1-62 E.1.5.2.3 Watershed Class .....                                                                                         E.1-62 i
E. 1-33 E.1.2.2.2 Timing of Release ........................
 
E.1-33E.1.2.2.3 Magnitude of Release ......................
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                        Pilgrim LR Proceeding SAMA Analysis                                                50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report ili ibl                                                                                            Operating License Renewal Stage E.1.5.2.4 Regional Economic Data -.............. ...................                                                     .
E.1-34 E.1.2.2.4 Release Category Bin Assignments
E.1.5.2.5 Agriculture Data .................... ...................                                                     .
....... .................
E.1.5.2.6 Meteorological Data ................                         ...................                             .
E.1-34 E.1.2.2.5 Mapping of Level 1 Results into the Various Release Categories
E.1.5.2.7 Emergency Response'Assumptions.                                                                 E.1-64        .
.E.1-35E.1.2.2.6 Collapsed Accident Progression Bins Source Terms
E.1.5.2.8 Core Inventory .....................                                                             1-64        .
.... ....... E.1-43 E.1.2.2.7 Release Magnitude Calculations
E.1.5.2.9 Source Terms ........................                                                         'E.1-66        .
..........................
E.1.5.3 Results ................................                                                               E.1-66        .
E.1-52 E.1.3 IPEEE Analysis ..........................
i  E.1.6 References      ...............................                                                             E.1-69        .
E.1-52 E.1.3.1 Seismic Analysis ..........................
E.2 EVALUATION OF 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        E.2.3 Final Screening and Cost Benefit Evaluation of SAMA Candidates (Phase II) E.2-2 E.2.4 Sensitivity Analyses ............................................ E.2-11 E.2.5 References .................                             .........................                         E.2-13 ii
E.1-52 E.1.3.2 Fire Analysis ....................... ....E.1-52 E.1.3.3 Other External Hazards ......................
 
E.1-54 E.1.4 PSA Model Peer Review and Difference between Current PSA Model and 1995 Update IPE .. .....................
Exhibit No. NRC000001 NRC - Applicant's Environmental Report          Pilgrim LR Proceeding SAMA Analysis                                  50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage     Ij List of Tables Table E.1-1 Core Damage Frequency Uncertainty .......................................                           E.1-2 Table E.1-2 PNPS PSA Model CDF Results by Major Initiators ........                                             E.1-3 Table E.1-3 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs ................                   E.1-4 Table E.1-4 Summary of PNPS PSA Core Damage Accident Class .........................                             E.1-28 Table E.1-5 Notation and Definitions for PNPS CET Functional Nodes Description .............                     E.1-29 Table E.1-7 PNPS Release Categories .....................                                                         E.1-35 Table E.1-6 Release Severity and Timing Classification Scheme Summary ...................                         E.1-35 Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States .......                           E.1-36 Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions ......................                       E.1 -44 Table E.1-10 Summary of PNPS Containment Event Tree Quantification ......................                         E.1-49 Table E.1-11 Collapsed Accident Progression Bin (CAPB) Source Terms .....................                         E.1-50 Table E.1-11 Collapsed Accident Progression Bin (CAPB) Source Terms (continued) ............................................................                             E. 1-51 Table E.1-12 PNPS Fire Updated Core Damage Frequency Results .........................                           E.1-53 Table E.1-13 Estimated Population Distribution within a 50-mile Radius .......................                     E.1-61 iii                                               &:
E.1-54 E.1.4.1 PSA Model Peer Review ........................
 
....E.1 -54 E.1.4.2 Major Differences between the Updated IPE PSA Model and 1995 Update IPE Model .................
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                        50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-14 PNPS Core Inventory (Becquerels) ........................................ E.1-65 Table E.1-15 Base Case Mean PDR and OECR Values ................................... E.1-67 Table E.1-16 Summary of Offsite Consequence Sensitivity Results .......................... E.1-68 Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation ..... E.2-15 Table E.2-2 Sensitivity Analysis Results..............................................E.245 iv
E.1-55 E.1.4.2.1 Core Damage
 
-Comparison to the PNPS 1995 Update IPE Model .........................
Exhibit No. NRC000001 NRC - Applicant's Environmental Report          Pilgrim LR Proceeding SAMA Analysis                                    50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage List of Figures Figure E.1-1 PNPS Radionuclide Release Category Summary ......................... E.1-31 Figure E.1-2 PNPS Plant Damage State Contribution to LERF ...........         .............. E.1-32 v0
E.1-55 E.1.4.2.2 Containment Performance
 
-Comparison to the Original PNPS IPE Model ..........
Exhibit No. NRC000001 NRC - Applicant's Environmental Report             Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage ATTACHMENT E.1 EVALUATION OF PSA MODEL
E.1-59 E.1.5 The MACCS2 Model -Level 3 Analysis ..............................
 
E.1-60 E.1.5.1 Introduction
Exhibit No. NRC000001 NRC - Applicant's Environmental Report          Pilgrim LR Proceeding SAMA Analysis                                  50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's 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 the MACCS2 code. The CAFTA code was used to develop the Pilgrim Nuclear Power Station (PNPS) PSA Level I and Level 2 models. This section provides the description of PNPS 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 I Analysis The PSA model (Level I and Level 2) used for the SAMA analysis was the most recent internal events risk model for PNPS (Revision 1, April 2003) [Reference E.1-1]. The PNPS PSA model and documentation has been updated to reflect the current plant operating configuration and design changes as of September 2001. The current PSA model reflects the accumulation of additional plant operating history and component failure and unavailability data as of December 2001. The PSA model also resolves all findings and observations during the industry peer review of the model, conducted in March 2000 [Reference E.1-1]. The PNPS model adopts the small event tree/ large fault tree approach and uses the CAFTA code for quantifying CDF. The Level I and Level 2 PNPS PSA analyses were originally developed and submitted to the NRC in September 1992 as the Pilgrim Nuclear Power Station Individual Plant Examination (IPE)
I ........ E.1-60 E.1.5.2 Input ................
Submittal [Reference E.1-2].
E.1-60 E.1.5.2.1 Projected Total Population by Spatial Element ..... ..........
The PSA model has been updated since the IPE due to the following.
E.1-61 E.1.5.2.2 Land Fraction ....................  
........ E.1-62 E.1.5.2.3 Watershed Class ..... E.1-62 i i Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage E.1.5.2.4 Regional Economic Data -. .............
E.1.5.2.5 Agriculture Data ....................
E.1.5.2.6 Meteorological Data ................E.1.5.2.7 Emergency Response'Assumptions.
E.1.5.2.8 Core Inventory
.....................E.1.5.2.9 Source Terms ........................
E.1.5.3 Results ................................E.1.6 References
...............................
................... .E.1-62................... .E.1-63................... .E.1-63...................
E.1-64...................
1-64...................
'E.1-66...................
E.1-66................
E.1-69 E.2 EVALUATION OF 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.2-11 E.2.5 References
.................
.........................
E.2-13 ii Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Ij List of Tables Table E.1-1 Core Damage Frequency Uncertainty  
.......................................
E.1-2 Table E.1-2 PNPS PSA Model CDF Results by Major Initiators  
........ E.1-3 Table E.1-3 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs ................
E.1-4 Table E.1-4 Summary of PNPS PSA Core Damage Accident Class .........................
E.1-28 Table E.1-5 Notation and Definitions for PNPS CET Functional Nodes Description  
.............
E.1-29 Table E.1-7 PNPS Release Categories  
.....................
E.1-35 Table E.1-6 Release Severity and Timing Classification Scheme Summary ...................
E.1-35 Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States ....... E.1-36 Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions  
......................
E.1 -44 Table E.1-10 Summary of PNPS Containment Event Tree Quantification  
......................
E.1-49 Table E.1-11 Collapsed Accident Progression Bin (CAPB)
Source Terms .....................
E.1-50 Table E.1-11 Collapsed Accident Progression Bin (CAPB) Source Terms (continued)  
............................................................
E. 1-51 Table E.1-12 PNPS Fire Updated Core Damage Frequency Results .........................
E.1-53 Table E.1-13 Estimated Population Distribution within a 50-mile Radius .......................
E.1-61 iii &:
Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-14 PNPS Core Inventory (Becquerels)  
........................................
E.1-65 Table E.1-15 Base Case Mean PDR and OECR Values ...................................
E.1-67 Table E.1-16 Summary of Offsite Consequence Sensitivity Results ..........................
E.1-68 Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation  
.....E.2-15 Table E.2-2 Sensitivity Analysis Results..............................................E.245 iv Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage List of Figures Figure E.1-1 PNPS Radionuclide Release Category Summary .........................
E.1-31 Figure E.1-2 PNPS Plant Damage State Contribution to LERF ...........  
..............
E.1-32 v0 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage ATTACHMENT E.1 EVALUATION OF PSA MODEL Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating 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 the MACCS2 code. The CAFTA code was used to develop the Pilgrim Nuclear Power Station (PNPS) PSA Level I and Level 2 models. This section providesthe description of PNPS 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 I AnalysisThe PSA model (Level I and Level
: 2) used for the SAMA analysis was the most recent internal events risk model for PNPS (Revision 1, April 2003)
[Reference E.1-1]. The PNPS PSA model and documentation has been updated to reflect the current plant operating configuration anddesign changes as of September 2001. The current PSA model reflects the accumulation of additional plant operating history and component failure and unavailability data as of December 2001. The PSA model also resolves all findings and observations during the industry peer review of the model, conducted in March 2000 [Reference E.1-1]. The PNPS model adopts the small event tree/ large fault tree approach and uses the CAFTA code for quantifying CDF. The Level I and Level 2 PNPS PSA analyses were originally developed and submitted to the NRC in September 1992 as the Pilgrim Nuclear Power Station Individual Plant Examination (IPE)Submittal
[Reference E.1-2].The PSA model has been updated since the IPE due to the following.
* In 1995, the original IPE model was changed in response to the NRC Request for Additional Information (RAI) received in April 1995 [Reference E.1-3]. Overall CDF was reduced from 5.85E-5/yr to 2.84E-5/yr. The reduction in CDF was due to removal of HPCI room cooling dependency, revised ADS success criteria, and improved HPCI/RCIC performance.
* In 1995, the original IPE model was changed in response to the NRC Request for Additional Information (RAI) received in April 1995 [Reference E.1-3]. Overall CDF was reduced from 5.85E-5/yr to 2.84E-5/yr. The reduction in CDF was due to removal of HPCI room cooling dependency, revised ADS success criteria, and improved HPCI/RCIC performance.
* Equipment performance
* Equipment performance - As data collection progresses, estimated failure rates and system unavailability data change.
-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 knowledgeand recommendations from internal and industry peer reviews.The PSA model contains the major initiators leading to core damage with baseline CDFs listed in Table E.1-2 [Reference E.1-1].The current PNPS PSA 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
* 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.
The PSA model contains the major initiators leading to core damage with baseline CDFs listed in Table E.1-2 [Reference E.1-1].
The current PNPS PSA 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% and E.1-1
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                      50-
E.1 -24
E.1 -24
:3 I9 J Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition SPC-MAI-MA-SPCA 3.01 E-03 1.005 Suppression pool This term represents RHR suppression pool cooling loop A cooling loop A out unavailable due to maintenance.
 
Phase I SAMAs to improve for maintenance availability and reliability of the RHR suppression pool cooling mode that have already been implemented include using drywell spray mode and fire protection cross-tie to provide redundantcontainment heat removal capability. Additional improvementswere evaluated in Phase II SAMAs 001 and 014.SPC-MAI-MA-SPCB 2.91E-03 1.005 Suppression pool This term represents RHR suppression pool cooling loop Bcooling loop B out unavailable due to maintenance.
Exhibit No. NRC000001
Phase I SAMAs to improve for maintenance availability and reliability of the RHR suppression pool cooling mode that have already been implemented include using drywell spray mode and fire protection cross-tie to provide redundant containment heat removal capability. Additional improvements were evaluated in Phase II SAMAs 001 and 014.DWS-MAI-MA-DWSA 3.18E-03 1.005 Drywell spray loop This term represents RHR drywell spray loop A unavailable due A out for to maintenance. Phase I SAMAs to improve availability and maintenance reliability of the RHR drywell spray mode that have already been implemented include using suppression pool cooling mode and fire protection cross-tie to provide redundant containment heatremoval capability.
:3                                                     I9 NRC - Applicant's Environmental Report SAMA Analysis                                                                                  J Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability   RRW     Event Description                                       Disposition SPC-MAI-MA-SPCA 3.01 E-03   1.005 Suppression pool           This term represents RHR suppression pool cooling loop A cooling loop A out         unavailable due to maintenance. Phase I SAMAs to improve for maintenance           availability and reliability of the RHR suppression pool cooling mode that have already been implemented include using drywell spray mode and fire protection cross-tie to provide redundant containment heat removal capability. Additional improvements were evaluated in Phase II SAMAs 001 and 014.
Additional improvements were evaluated in Phase II SAMA 009.E.1-25
SPC-MAI-MA-SPCB 2.91E-03     1.005 Suppression pool           This term represents RHR suppression pool cooling loop B cooling loop B out         unavailable due to maintenance. Phase I SAMAs to improve for maintenance           availability and reliability of the RHR suppression pool cooling mode that have already been implemented include using drywell spray mode and fire protection cross-tie to provide redundant containment heat removal capability. Additional improvements were evaluated in Phase II SAMAs 001 and 014.
_A , Yc-r-i 01*Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Table E.1-3 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition ADS-XHE-FO-XIS2 1 .45E-03 1.005 Operator fails to This term represents operator failure to manually open the SRVs perform emergency for depressurization during a small LOCA. Phase I SAMAs, depressurization including improvement of procedures and installation of during small LOCA instrumentation to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented.
DWS-MAI-MA-DWSA 3.18E-03     1.005 Drywell spray loop         This term represents RHR drywell spray loop A unavailable due A out for                 to maintenance. Phase I SAMAs to improve availability and maintenance               reliability of the RHR drywell spray mode that have already been implemented include using suppression pool cooling mode and fire protection cross-tie to provide redundant containment heat removal capability. Additional improvements were evaluated in Phase II SAMA 009.
No additional Phase II SAMAs were recommended for this subject.E.1-26 Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage E.1.2 PSA Model -Level 2 Analysis E.1.2.1 Containment Performance Analysis The PNPS Level 2 PSA model used for the SAMA analysis is the most recent internal events risk model, which is an updated version of the model used in the IPE [References E.1-2 and E.1-3].The Level 2 PSA model used for the SAMA analysis, Revision 1, reflects the PNPS operating configuration and design changes as of September 2001. Specifically, the Level 2 model hasbeen updated to incorporate insights from the independent BWROG peer review.The PNPS Level 2 model includes two types of considerations:
E.1-25
(1) a deterministic analysis of thephysical processes for a spectrum of severe accident progressions, and (2) a probabilistic analysis component in which the likelihood of the various outcomes are assessed.
 
Thedeterministic analysis examines the response of the containment to the physical processes during a severe accident.
Exhibit No. NRC000001 Yc-_A
This response is performed by* utilization of the MAAP code [Reference E. 14] 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.
,                                     NRC - Applicant's Environmental Report SAMA Analysis r-i                                                                       01*
Examples include debris coolability, pressure spikes due to ex-vessel steam explosions, scoping calculation of direct containmentheating, 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.
Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs Event Name Probability   RRW     Event Description                                   Disposition ADS-XHE-FO-XIS2 1.45E-03     1.005 Operator fails to         This term represents operator failure to manually open the SRVs perform emergency         for depressurization during a small LOCA. Phase I SAMAs, depressurization           including improvement of procedures and installation of during small LOCA         instrumentation to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented. No additional Phase IISAMAs were recommended for this subject.
The Level 2 analysis examined the dominant accident sequences and the resulting plant damage states (PDS) defined in Level 1. The Level I analysis involves the assessment of those scenarios that could lead to core damage.
E.1-26
A list of the PDS groups and descriptions from the Level 2 analysis is presented in Table E.1-4.
 
A full Level 2 model was developed for the IPE and completed at the same time as the Level 1 model. The Level 2 model consists of a single containment event tree (CET) 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.1-5.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 nntontfil fnr incriv haIth affontc tn theg nohlh- Tha frong icnev role infin *nnrnyimqtcmlv Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-4 Summary of PNPS PSA Core Damage Accident Class PoS S l Point % of Total Group SipiidDescription Estimate CDF LOCAs Large and small break LOCA with initial or long-term loss 1.16E-7 1.80 of core cooling.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report          Pilgrim LR Proceeding SAMA Analysis                                  50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.2 PSA Model - Level 2 Analysis E.1.2.1 Containment Performance Analysis The PNPS Level 2 PSA model used for the SAMA analysis is the most recent internal events risk model, which is an updated version of the model used in the IPE [References E.1-2 and E.1-3].
Core damage results at low or high reactor pressure.
The Level 2 PSA model used for the SAMA analysis, Revision 1, reflects the PNPS operating configuration and design changes as of September 2001. Specifically, the Level 2 model has been updated to incorporate insights from the independent BWROG peer review.
For most PDS, late injection and containment heat removal are available.
The PNPS 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
TRANS Short and long-term transient events. Core damage 2.43E-7 3.79 results at either low or high reactor pressure.
* utilization of the MAAP code [Reference E. 14] to simulate severe accidents that have been identified as dominant contributors to core damage in the Level 1 analysis, and
Late injection and containment heat removal are available.
* 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.
SBO SBO involving a loss of high-pressure injection. Core 1.48E-7 2.31damage results at either low (stuck-open SRV) or high reactor pressure. All accident mitigating functions are recoverable when AC power is restored.VSLRUPT Vessel rupture event resulting In LOCA beyond ECCS 4.OOE-9 0.06 capability.
The Level 2 analysis examined the dominant accident sequences and the resulting plant damage states (PDS) defined in Level 1. The Level I analysis involves the assessment of those scenarios that could lead to core damage. A list of the PDS groups and descriptions from the Level 2 analysis is presented in Table E.1-4.
All PDS result in core damage at low reactor pressure with late injection available.
A full Level 2 model was developed for the IPE and completed at the same time as the Level 1 model. The Level 2 model consists of a single containment event tree (CET) 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.1-5.
ATWS Short-term ATWS that leads to early core damage at high 3.39E-8 0.53reactor pressure following loss of reactivity control and rapid containment pressurization. Reactor coolant system leakage rates associated with boil-off of coolantthrough the cycling of SRVs/SV with early core melt subsequent to containment overpressure failure. Late injection and containment heat removal are available.
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 nntontfil fnr incriv haIth affontc tn theg nohlh- Tha frong icnev role infin*nnrnyimqtcmlv
ISLOCA Large and small break interfacing system LOCA outside 4.00E-9 0.06 containment.
 
Core damage results at low or high reactor pressure with a bypassed containment.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                Pilgrim LR Proceeding SAMA Analysis                                        50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report kmr'                                                                            Operating License Renewal Stage Table E.1-4 Summary of PNPS PSA Core Damage Accident Class PoS                     S                                             l     Point         % of Total Group                     SipiidDescription                               Estimate             CDF LOCAs Large and small break LOCA with initial or long-term loss             1.16E-7             1.80 of core cooling. Core damage results at low or high reactor pressure. For most PDS, late injection and containment heat removal are available.
TW Containment decay heat removal systems are not 5.86E-6 91.45 available and coolant recirculation to the torus over pressurizes the containment to failure or venting. The torus is saturated.
TRANS   Short and long-term transient events. Core damage                     2.43E-7             3.79 results at either low or high reactor pressure. Late injection and containment heat removal are available.
Total 6.41 E-06 1.OOE+00 E.1-28 Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Table E.1-5 Notation and Definitions for PNPS CET Functional Nodes Description CET Node CET Functional Node Description Plant Damage State This top event represents the initiators considered in the containment Event (PDSEVNT) performance analysis.RPV Pressure at This top event identifies the status of the reactor pressure vessel (RPV)Vessel Failure pressure.
SBO     SBO involving a loss of high-pressure injection. Core                 1.48E-7             2.31 damage results at either low (stuck-open SRV) or high reactor pressure. All accident mitigating functions are recoverable when AC power is restored.
RPV@VF is set to success when RPV pressure is low.(RPV@VF) RPV@VF is set to failure when RPV is high.In-Vessel Cooling This top event addresses the recovery of coolant injection into the vessel Recovery (IN-REC) after core degradation, but prior to vessel breach. This top event considers the possibility of low-pressure injection systems working once the RPV is depressurized.
VSLRUPT Vessel rupture event resulting In LOCA beyond ECCS                   4.OOE-9             0.06 capability. All PDS result in core damage at low reactor pressure with late injection available.
Vessel Failure (VF) This top event addresses recovery from core degradation within the vessel and the prevention of vessel head thermal attack. Core melt recovery requires the recovery of core cooling prior to core blocking or relocation of molten debris to the lower plenum and thermal attack of the vessel head.Early Containment This top event node considers the potential loss of containment integrity at, Failure (CFE) or before, vessel failure.
ATWS     Short-term ATWS that leads to early core damage at high               3.39E-8             0.53 reactor pressure following loss of reactivity control and rapid containment pressurization. Reactor coolant system leakage rates associated with boil-off of coolant through the cycling of SRVs/SV with early core melt subsequent to containment overpressure failure. Late injection and containment heat removal are available.
Several phenomena are considered credible mechanisms for early containment failure.
ISLOCA   Large and small break interfacing system LOCA outside                 4.00E-9             0.06 containment. Core damage results at low or high reactor pressure with a bypassed containment.
They may occur alone or in combination.
TW       Containment decay heat removal systems are not                       5.86E-6           91.45 available and coolant recirculation to the torus over pressurizes the containment to failure or venting. The torus is saturated.
The phenomena are containment isolation failure; containment bypass; containment overpressure failure at vessel breach;hydrogen deflagration or detonation; fuel-coolant interactions (steam explosions);
Total   6.41 E-06       1.OOE+00 E.1-28
high pressure melt ejection and subsequent direct containment heating; and drywell steel shell melt-through.I Early Release to Torus (EPOOL)This top event node considers the importance of early torus pool scrubbing in mitigating the magnitude of fission products released from the damaged core. Success implies that fission product transport path subsequent toearly containment failure is through the torus water and the torus airspace.That is, the torus pool is not bypassed.
 
Failure involves a release into the drywell.Debris Cooled Ex-vessel (DCOOL)This top event considers the delivery of water to the drywell, via drywell sprays, or via injection to the RPV and drainage out an RPV breach onto the drywell floor. Success implies the availability of water and the formation of a coolable debris bed such that concrete attack is precluded.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report            Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage   ( I Table E.1-5 Notation and Definitions for PNPS CET Functional Nodes Description CET Node                               CET Functional Node Description Plant Damage State   This top event represents the initiators considered in the containment Event (PDSEVNT)       performance analysis.
Failure implies that the molten core attacks concrete in the reactor pedestal, that core debris remains hot, and sparing of the concrete decomposition products through the melt releases the less volatile fission products to the containment atmosphere.
RPV Pressure at       This top event identifies the status of the reactor pressure vessel (RPV)
E.1-29 J, Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-5 Notation and Definitions for PNPS CET Functional Nodes Description (Continued)
Vessel Failure       pressure. RPV@VF is set to success when RPV pressure is low.
CET Node CET Functional Node Description Late Containment This top event addresses the potential loss of containment integrity in the Failure (CFL) long-term.
(RPV@VF)             RPV@VF isset to failure when RPV is high.
Late containment failure may result from long-term steam and non-condensable gas generation from the attack of molten core debris on concrete.Late Release to This top event node considers the importance of late torus pool scrubbing in Torus (LPOOL) mitigating the magnitude of fission products released from the damaged core. Success implies that fission product transport path subsequent to latecontainment failure is through the torus water and the torus airspace. That is, the torus pool is not bypassed. Failure involves a release into the drywell.Fission Product This top event addresses fission product releases from the fuel into the Removal (FPR) containment and airborne fission product removal mechanisms within thecontainment structure to characterize potential magnitude of fission product releases to the environment should the containment fail. Failure impliesthat most of the fission products from the fuel and containment are ultimately released to the environment without mitigation.Reactor Building This top event is used to assess the ability of the reactor building to retain (RB) fission products released from containment.
In-Vessel Cooling     This top event addresses the recovery of coolant injection into the vessel Recovery (IN-REC)     after core degradation, but prior to vessel breach. This top event considers the possibility of low-pressure injection systems working once the RPV is depressurized.
Success of top event RB is defined to be a reduction of the containment release magnitude.
Vessel Failure (VF)   This top event addresses recovery from core degradation within the vessel and the prevention of vessel head thermal attack. Core melt recovery requires the recovery of core cooling prior to core blocking or relocation of molten debris to the lower plenum and thermal attack of the vessel head.
E.1-30  
Early Containment     This top event node considers the potential loss of containment integrity at, Failure (CFE)         or before, vessel failure. Several phenomena are considered credible mechanisms for early containment failure. They may occur alone or in combination. The phenomena are containment isolation failure; containment bypass; containment overpressure failure at vessel breach; hydrogen deflagration or detonation; fuel-coolant interactions (steam explosions); high pressure melt ejection and subsequent direct containment heating; and drywell steel shell melt-through.
.Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage fW- 'Lat, Eary LowFRlease  
Early Release to     This top event node considers the importance of early torus pool scrubbing Torus (EPOOL)        in mitigating the magnitude of fission products released from the damaged core. Success implies that fission product transport path subsequent to early containment failure is through the torus water and the torus airspace.
,.0.52% a e Hgh ReeaseO4. OO/o a\Jum Release-'/uly H Eal/Ah Release.76%Ealy Medun Rlease 1 mo 1°/Late MWc 2A AN-No Con~imnat Failure 1.73%LateL R.lease 70.65%Figure E.1-1 PNPS Radionuclide Release Category Summary E.1-31 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Transierts Aticipated Transiert wthot 201%Sacam 39.82%1. .Interfaang System LOCAs LOCus t 13 2 YO 0.01%Ve sa Rome 0.01%Staficn Blackout 57.03%Figure E.1-2 PNPS Plant Damage State Contribution to LERF E.1-32 Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage E.1.2.2 Radionuclide Analysis E.1.2.2.1 IntroductionA 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.
That is, the torus pool is not bypassed. Failure involves a release into the drywell.
These end points represent complete severe accident sequences from initiating event to release of radionuclides to the environment.
Debris Cooled Ex-     This top event considers the delivery of water to the drywell, via drywell vessel (DCOOL)        sprays, or via injection to the RPV and drainage out an RPV breach onto the drywell floor. Success implies the availability of water and the formation of a coolable debris bed such that concrete attack is precluded. Failure implies that the molten core attacks concrete in the reactor pedestal, that core debris remains hot, and sparing of the concrete decomposition products through the melt releases the less volatile fission products to the containment atmosphere.
The Level I 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 sequencesproduced 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.
E.1-29 J,
The predicted source term associated witheach release category, including both the timing and magnitude of the release, is determined using the results of MAAP calculations
 
[Reference E.1-4].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. PNPS characterizes the release timing relative to the time at which the release begins, measured from the time of accident initiation.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-5 Notation and Definitions for PNPS CET Functional Nodes Description (Continued)
Two timing categories are used: early (0-24 hours) and late (>24 hours).Based on MAAP calculations for a spectrum of severe accident sequences, PNPS expects that an Emergency Action Level (as defined by the PNPS Emergency Plan) will be reached within the first half hour after accident initiation.
CET Node                             CET Functional Node Description Late Containment       This top event addresses the potential loss of containment integrity in the Failure (CFL)         long-term. Late containment failure may result from long-term steam and non-condensable gas generation from the attack of molten core debris on concrete.
Reaching an Emergency Action Level initiates a formaldecision-making process that is designed to provide public protective actions. Within 24 hours of accident initiation, the Level 2 analysis assumed that off-site protective measures would be effective. Therefore, the definitions of the release timing categories are as follows.* Early releases are CET end-states involving containment failure prior to or at vessel failure or after vessel failure and occurring within 0 to 24 hours measured from the time of accident initiation and for which minimal offsite protective measures would be accomplished.
Late Release to       This top event node considers the importance of late torus pool scrubbing in Torus (LPOOL)           mitigating the magnitude of fission products released from the damaged core. Success implies that fission product transport path subsequent to late containment failure is through the torus water and the torus airspace. That is, the torus pool is not bypassed. Failure involves a release into the drywell.
Fission Product       This top event addresses fission product releases from the fuel into the Removal (FPR)         containment and airborne fission product removal mechanisms within the containment structure to characterize potential magnitude of fission product releases to the environment should the containment fail. Failure implies that most of the fission products from the fuel and containment are ultimately released to the environment without mitigation.
Reactor Building       This top event is used to assess the ability of the reactor building to retain (RB)                   fission products released from containment. Success of top event RB is defined to be a reduction of the containment release magnitude.
E.1-30
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                    Pilgrim LR Proceeding SAMA Analysis                                              50-293-LR, 06-848-02-LR
                                                                          . Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage   fW- '
Ealy Medun Rlease Eary LowFRlease       ,              1 mo1°/
                      .0.52%       a                 '/uly          HAh Release
                                                                      .76%
Lat,eHgh Reease O4.
OO/o a\
Eal/  -No Con~imnat Failure Late MWc Jum Release-                                                           1.73%
2A AN LateL   R.lease 70.65%
Figure E.1-1 PNPS Radionuclide Release Category Summary E.1-31
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                     Pilgrim LR Proceeding SAMA Analysis                                              50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Transierts                 Interfaang System LOCAs    LOCus Aticipated Transiert wthot 201%                                 2 t 13YO          0.01%
Sacam 39.82%                                                                   Vesa Rome 0.01%
: 1.               .
Staficn Blackout 57.03%
Figure E.1-2 PNPS Plant Damage State Contribution to LERF E.1-32
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report          Pilgrim LR Proceeding SAMA Analysis                                  50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's 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 I 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 [Reference E.1-4].
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. PNPS characterizes the release timing relative to the time at which the release begins, measured from the time of accident initiation. Two timing categories are used: early (0-24 hours) and late (>24 hours).
Based on MAAP calculations for a spectrum of severe accident sequences, PNPS expects that an Emergency Action Level (as defined by the PNPS Emergency Plan) will be reached within the first half hour after accident initiation. Reaching an Emergency Action Level initiates a formal decision-making process that is designed to provide public protective actions. Within 24 hours of accident initiation, the Level 2 analysis assumed that off-site protective measures would be effective. Therefore, the definitions of the release timing categories are as follows.
* Early releases are CET end-states involving containment failure prior to or at vessel failure or after vessel failure and occurring within 0 to 24 hours measured from the time of accident initiation and for which minimal offsite protective measures would be accomplished.
* Late releases are CET end-states involving containment failure greater than 24 hours from the time of accident initiation, for which offsite measures are fully effective.
* Late releases are CET end-states involving containment failure greater than 24 hours from the time of accident initiation, for which offsite measures are fully effective.
E.1-33 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 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.1-33
[Reference E.1-5]. The CsI release 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 four distinct radionuclide release categories or bins according to release magnitude as follows:
 
(1) High (HI) -A radionuclide release of sufficient magnitude to have the potential to cause early fatalities.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                        50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 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 [Reference E.1-5]. The CsI release fraction indicates the fraction of in-vessel radionuclides escaping to the environment.
This implies a total integrated release of >10% of the initial core inventory of Csl [Reference E.1-5].1 (2) Medium (MED) -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 [Reference E. 1-5].2 (3) Low (LO) -A radionuclide release with the potential for latent health effects. This implies a total integrated release of between 0.001% an'd 1% of the initial core inventory of CsI.(4) Negligible (NCF) -A radionuclide release that is less than or equal to the containment design base leakage. This implies total integrated release of<0.001% of the initial core inventory of Csl.The "total integrated release" as used in the above categories is defined as the integrated release within 36 hours after RPV failure. If no RPV failure occurs, then the "total integrated release" is defined as the integrated release within 36 hours after accident initiation.
(Noble gas release'levels are non-informative since release of the total core inventory of noble gases is essentially complete given containment failure).
E.1.2.2.4 Release Category Bin Assignments Table E.1-6 summarizes the scheme used to bin sequences with respect to magnitude of release, based on the predicted Csl release fraction and release timing. The combi nation of release magnitude and timing produce seven distinct release categories for source terms. These are the representative release categories presented in Table E. 1-7.1. Once the Csl source term exceeds 0.1, the source term Is large enough that doses above the early fatality threshold can sometimes occur within a population center a few miles from the site.2. The reference document indicates that for'Csl release fractions of 1 to 10%, the number oflatent fatalities is found to be at least 10% of the latent fatalities for the highest release.E.1-34 Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Table E.1-6 Release Severity and Timing Classification Scheme Summary Release Severity Release Timing Classification Classification Time of Initial Release from Category Csl % Release Category Accident Initiation High Greater than 10 Early (E) Less than 24 hours Medium i to 10 Low 0.001 to 1Late (L) Greater than 24 hours Negligible Less than 0.001 Table E.1-7 PNPS Release Categories Timing of Magnitude of Release Release Low Medium High Early Early/Low Early/Med Early/High NCF Late Late/Low Late/Med Late/High)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 damageaccident sequences and fission product release categories).
The source terms were grouped into four distinct radionuclide release categories or bins according to release magnitude as follows:
In the PDS analysis, Level 1 resultswere grouped
(1) High (HI) - 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 Csl [Reference E.1-5].1 (2)   Medium (MED) - 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 [Reference E. 1-5].2 (3)   Low (LO) - A radionuclide release with the potential for latent health effects. This implies a total integrated release of between 0.001% an'd 1% of the initial core inventory of CsI.
("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 systemsimportant 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.
(4)   Negligible (NCF) - A radionuclide release that is less than or equal to the containment design base leakage. This implies total integrated release of
A PDStherefore 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.
            <0.001% of the initial core inventory of Csl.
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, accidentprevention systems, and mitigation systems that have been noted in the Level 1 analysis are carried into the Level 2 analysis.
The "total integrated release" as used in the above categories is defined as the integrated release within 36 hours after RPV failure. If no RPV failure occurs, then the "total integrated release" is defined as the integrated release within 36 hours after accident initiation.
In this latter analysis, the repair or recovery of failed equipment is not allowed unless an explicit evaluation, including a consideration of E.1-35 Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage 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 Ianalysis to the CET.* Containment status. The containment status is explicitly transferred from the Level 1analysis to the CET. This includes recognition of whether the containment is bypassed or is intact at the onset of core damage.* Accident sequence timing. Differences in accident sequence timing are transferred withthe Level 1 sequences.
E.1.2.2.4     Release Category Bin Assignments Table E.1-6 summarizes the scheme used to bin sequences with respect to magnitude of release, based on the predicted Csl release fraction and release timing. The combi nation of release magnitude and timing produce seven distinct release categories for source terms. These are the representative release categories presented in Table E. 1-7.
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 48 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.
: 1. Once the Csl source term exceeds 0.1, the source term Is large enough that doses above the early fatality threshold can sometimes occur within a population center a few miles from the site.
Table E.1-8 provides a description of the PNPS PDS that are used to summarize the Level 1 results."ms Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States Point %fD PDS Description Estimate %ofCDF PDS-1 Long-term LOCA with loss of high-pressure core makeup O.OOE+00 0.00 from HPCI and RCIC, loss of containment heat removal, and failure to depressurize the primary system for low-pressure core makeup.
: 2. The reference document indicates that for'Csl release fractions of 1 to 10%, the number of latent fatalities is found to be at least 10% of the latent fatalities for the highest release.
Core damage results at highprimary system pressure.
E.1-34
Late injection from low-pressure systems (core spray, LPCI, and firewater) is available,provided primary system depressurization occurs.
 
The containment is vented and intact.PDS-2 Long-term LOCA with loss of both high-pressure core 1.05E-11 <0.001 makeup (HPCI and RCIC) and containment heat removal.Core damage results at high primary system pressure.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                  Pilgrim LR Proceeding SAMA Analysis                                          50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-6 Release Severity and Timing Classification Scheme Summary Release Severity                                         Release Timing Classification                                       Classification       Time of Initial Release from Category             Csl %Release                   Category               Accident Initiation High               Greater than 10 Early (E)             Less than 24 hours Medium                   i to 10 Low                   0.001 to 1 Late (L)             Greater than 24 hours Negligible           Less than 0.001 Table E.1-7 PNPS Release Categories Timing of                             Magnitude of Release Release             Low               Medium               High Early         Early/Low             Early/Med           Early/High           NCF Late           Late/Low             Late/Med             Late/High                               C )
Because containment venting fails, containment failure occurs long-term.
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.
Late injection is available from low-pressure systems (core spray, LPCI, and fire water)provided they survive containment failure.E.1-36 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
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.
Point PDS Description Estimate % of CDFPDS-3 Short-term LOCA with loss of high-pressure core makeup, 8.68E-08 1.35 and failure to depressurize the primary system for low-pressure core makeup. Core damage occurs at highprimary system pressure.
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 E.1-35
Late injection from core spray, LPCI, and firewater is available, provided primary system depressurization occurs. Containment heat removal is available.PDS-4 Short-term LOCA with loss of high-pressure core makeup, O.OOE+00 <0.001 loss of containment heat removal, and failure to depressurize the primary system for low-pressure core makeup. Core damage occurs at high primary system pressure.
 
Late injection from core spray, LPCI, and firewater is available, provided primary systemdepressurization occurs.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage adverse environments where appropriate, has been performed as part of the Level 2 analysis.
Unlike PDS-3, containment heat removal is unavailable.
* RPV status. The RPV pressure condition is explicitly transferred from the Level I analysis to the CET.
PDS-5 Long-term LOCA with loss of high-pressure core makeup 0.OOE+00 0.00 and containment heat removal. Core damage occurs atlow primary system.
* 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.
Late injection is available from low-pressure systems (core spray, LPCI, and fire water). The containment is vented and intact.PDS-6 Long-term large LOCA. High-pressure core makeup from 0.00E+00 0.00 HPCI and RCIC are unavailable due to the large LOCA.Because containment venting fails, containment failure occurs long-term. Late injection is available from low-pressure systems (core spray, LPCI, and fire water)provided they survive containment failure. Core damage occurs at low primary system pressure.PDS-7 Short-term large LOCA with loss of core cooling. Core 1.12E-09 0.08damage results at low primary system pressure.
* Accident sequence timing. Differences in accident sequence timing are transferred with the Level 1 sequences. Timing affects such sequences as SBO, internal flooding, and containment bypass (ISLOCA).
Late injection from firewater cross tie and containment heat removal are available.
This transfer of information allows timing to be properly assessed in the Level 2 analysis.
PDS0- Short-term large LOCA with loss of core cooling. Core 4.43E-09 0.07damage results at low primary system pressure.
Based on the above criteria, the Level 1 results were binned into 48 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.1-8 provides a description of the PNPS PDS that are used to summarize the Level 1 results.
Late injection from firewater cross tie is available.
"ms                                               Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States PDS                              Description                                  Point Estimate          %fD
However, unlike PDS-7, containment heat removal is unavailable.
                                                                                                      %ofCDF PDS-1       Long-term LOCA with loss of high-pressure core makeup           O.OOE+00           0.00 from HPCI and RCIC, loss of containment heat removal, and failure to depressurize the primary system for low-pressure core makeup. Core damage results at high primary system pressure. Late injection from low-pressure systems (core spray, LPCI, and firewater) is available, provided primary system depressurization occurs. The containment is vented and intact.
Q.E.1-37 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
PDS-2       Long-term LOCA with loss of both high-pressure core             1.05E-11           <0.001 makeup (HPCI and RCIC) and containment heat removal.
Point PDS Description Estimate l of CDF PDS-9 Short-term LOCA with loss of high and low-pressure core 3.64E-09 0.06%cooling. Because the primary system is depressurized, core damage results at low primary system pressure. Late injection from SSW system, containment venting, andcontainment heat removal are available.
Core damage results at high primary system pressure.
PDS-10 Short-term LOCA with loss of high and low-pressure core O.OOE+00 0.00 cooling. Because the primary system is depressurized, core damage results at low primary system pressure.
Because containment venting fails, containment failure occurs long-term. Late injection is available from low-pressure systems (core spray, LPCI, and fire water) provided they survive containment failure.
Late injection from SSW system and containment heat removal are available.
E.1-36
However, unlike PDS-9, containment ventingis not available.
 
PDS-11 Short-term LOCA with loss of high and low-pressure core O.OOE+00 0.00 cooling. Core damage results at low primary system pressure.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report           Pilgrim LR Proceeding SAMA Analysis                                    50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
Late injection from SSW system is available.
Point PDS                               Description                           Estimate           %of CDF PDS-3     Short-term LOCA with loss of high-pressure core makeup,       8.68E-08             1.35 and failure to depressurize the primary system for low-pressure core makeup. Core damage occurs at high primary system pressure. Late injection from core spray, LPCI, and firewater is available, provided primary system depressurization occurs. Containment heat removal is available.
However, unlike PDS-9, containment venting andcontainment heat removal are unavailable.
PDS-4     Short-term LOCA with loss of high-pressure core makeup,       O.OOE+00           <0.001 loss of containment heat removal, and failure to depressurize the primary system for low-pressure core makeup. Core damage occurs at high primary system pressure. Late injection from core spray, LPCI, and firewater is available, provided primary system depressurization occurs. Unlike PDS-3, containment heat removal is unavailable.
PDS-12 Transient with a loss of long-term decay heat removal.
PDS-5     Long-term LOCA with loss of high-pressure core makeup and containment heat removal. Core damage occurs at low primary system. Late injection is available from low-0.OOE+00            0.00 Q.
Core 2.37E-08 0.37damage results at high primary system pressure.
pressure systems (core spray, LPCI, and fire water). The containment is vented and intact.
Late in-vessel and ex-vessel injection is available. The containment is vented and remains intact at the time ofcore damage.
PDS-6     Long-term large LOCA. High-pressure core makeup from           0.00E+00           0.00 HPCI and RCIC are unavailable due to the large LOCA.
PDS-13 Transient with a loss of long-term decay heat removal. Core 3.75E-06 58.5 damage results at high primary system pressure.
Because containment venting fails, containment failure occurs long-term. Late injection is available from low-pressure systems (core spray, LPCI, and fire water) provided they survive containment failure. Core damage occurs at low primary system pressure.
Late in-vessel and ex-vessel injection is available.
PDS-7     Short-term large LOCA with loss of core cooling. Core         1.12E-09           0.08 damage results at low primary system pressure. Late injection from firewater cross tie and containment heat removal are available.
Unlike PDS-12 containment venting fails.PDS-14 Short-term transient with failure to depressurize the primary 1 .52E-07 2.37 system. Core damage results at high primary systempressure. Late in-vessel and ex-vessel injection is available.
PDS0-     Short-term large LOCA with loss of core cooling. Core         4.43E-09           0.07 damage results at low primary system pressure. Late injection from firewater cross tie is available. However, unlike PDS-7, containment heat removal is unavailable.
Containment heat removal from RHR is available.
E.1-37
PDS Short-term transient with failure to depressurize the primary 5.07E-08 0.79 system. Core damage results at high primary system pressure.
 
Late in-vessel and ex-vessel injection is available.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
Containment heat removal from RHR is available.
Point PDS                             Description                               Estimate       l   of CDF PDS-9   Short-term LOCA with loss of high and low-pressure core         3.64E-09             0.06%
cooling. Because the primary system is depressurized, core damage results at low primary system pressure. Late injection from SSW system, containment venting, and containment heat removal are available.
PDS-10   Short-term LOCA with loss of high and low-pressure core         O.OOE+00           0.00 cooling. Because the primary system is depressurized, core damage results at low primary system pressure. Late injection from SSW system and containment heat removal are available. However, unlike PDS-9, containment venting is not available.
PDS-11   Short-term LOCA with loss of high and low-pressure core         O.OOE+00           0.00 cooling. Core damage results at low primary system pressure. Late injection from SSW system is available.
However, unlike PDS-9, containment venting and containment heat removal are unavailable.
PDS-12   Transient with a loss of long-term decay heat removal. Core     2.37E-08           0.37 damage results at high primary system pressure. Late in-vessel and ex-vessel injection is available. The containment is vented and remains intact at the time of core damage.
PDS-13   Transient with a loss of long-term decay heat removal. Core       3.75E-06           58.5 damage results at high primary system pressure. Late in-vessel and ex-vessel injection is available. Unlike PDS-12 containment venting fails.
PDS-14   Short-term transient with failure to depressurize the primary     1.52E-07           2.37 system. Core damage results at high primary system pressure. Late in-vessel and ex-vessel injection is available. Containment heat removal from RHR is available.
PDS Short-term transient with failure to depressurize the primary     5.07E-08           0.79 system. Core damage results at high primary system pressure. Late in-vessel and ex-vessel injection is available. Containment heat removal from RHR is available. However, containment venting is not available.
E.1-38
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report            Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report I
Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
PDS                            Description                                Point            % of CDF PDS-16  Short-term transient with failure to depressurize the primary  4.89E-09            0.08 system. Core damage results at high primary system pressure. Late in-vessel and ex-vessel injection is available. Containment heat removal from RHR is not available, but containment venting is available.
PDS-17  Short-term transient with failure to depressurize the primary  2.53E-09            0.04 system. Core damage results at high primary system pressure. Late in-vessel and ex-vessel injection is available. Neither containment heat removal from RHR nor containment venting is available.
PDS-18  Transient with a loss of long-term decay heat removal.          1.56E-06            24.40 Core damage results at low primary system pressure. Late in-vessel and ex-vessel injection is available. The containment is vented and remains intact at the time of core damage.
PDS-19  Transient with a loss of long-term decay heat removal.          5.24E-07            8.18        Cl Core damage results at low primary system pressure. Late in-vessel and ex-vessel injection is available. Unlike PDS-18 containment venting fails.
PDS-20  Long-term transients with loss of core cooling. Core            6.78E-11            0.001 damage results at low primary system pressure. No late injection, but containment heat removal is available.
PDS-21  Short-term transients (IORV) with loss of core cooling.        8.18E-09            0.13 Core damage results at low primary system pressure. Late injection and containment heat removal are available.
PDS-22  Short-term transients with loss of core cooling. Core          1.08E-09            0.02 damage results at low primary system pressure. Late injection and containment heat removal are available.
However, containment venting is not available.
However, containment venting is not available.
E.1-38 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
PDS-231 Short-term transients with loss of core cooling. Core           O.OOE+00             0.00 damage results at low primary system pressure. Late injection and containment venting are available, but containment heat removal is not available.
PDS Description Point % of CDF PDS-16 Short-term transient with failure to depressurize the primary 4.89E-09 0.08 system. Core damage results at high primary system pressure.
PDS-24 Similar to PDS-23, except that containment venting is not       4.98E-09             0.08 available.
Late in-vessel and ex-vessel injection is available.
E.1-39
Containment heat removal from RHR is not available, but containment venting is available.PDS-17 Short-term transient with failure to depressurize the primary 2.53E-09 0.04 system. Core damage results at high primary systempressure. Late in-vessel and ex-vessel injection isavailable. Neither containment heat removal from RHR norcontainment venting is available.
 
PDS-18 Transient with a loss of long-term decay heat removal. 1 .56E-06 24.40Core damage results at low primary system pressure.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
Late in-vessel and ex-vessel injection is available.
Point PDS                             Description                               Estimate         C of CDF PDS-25   Short-term transients with loss of core cooling. Core           2.57E-09           0.04 damage results at low primary system pressure. No late injection, but containment heat removal and containment venting are available.
The containment is vented and remains intact at the time of core damage.PDS-19 Transient with a loss of long-term decay heat removal. 5.24E-07 8.18Core damage results at low primary system pressure. Late in-vessel and ex-vessel injection is available.
PDS-26   Similar to PDS-25, except that containment venting is not       1.24E-08           0.19 available.
Unlike PDS-18 containment venting fails.PDS-20 Long-term transients with loss of core cooling. Core 6.78E-11 0.001 damage results at low primary system pressure.
PDS-27   Short-term transients with loss of core cooling. Core           4.40E-11           0.001 damage results at low primary system pressure. Late injection and containment heat removal are not available.
No late injection, but containment heat removal is available.
However, containment venting is available PDS-28   Short-term transients with loss of core cooling. Core           1.10E-09           0.02 damage results at low primary system pressure. Late injection, containment heat removal and containment venting are not available.
PDS-21 Short-term transients (IORV) with loss of core cooling.
PDS-29   Long-term SBO involving loss of injection at high primary       1.41 E-07           2.21 system pressure from battery depletion. All accident-mitigating functions are recoverable when AC power is restored.
8.18E-09 0.13Core damage results at low primary system pressure. Late injection and containment heat removal are available.
PDS-30   Short-term SBO sequence involving a loss of high-pressure       O.OOE+00           0.00 injection at high primary system pressure from loss of all AC power and DC power or failure of SRVs. All accident-mitigating functions are recoverable when offsite power is restored.
PDS-22 Short-term transients with loss of core cooling.
PDS-31 Long-term SBO sequence Involving a loss of high-pressure         2.60E-09           0.04 injection due to one stuck-open safety relief valve or long-term failure of HPCI and RCIC and subsequent failure to depressurize the primary system. Core damage results at low primary system pressure. All accident-rnitigating functions are recoverable when offsite power isrestored.
Core 1.08E-09 0.02 damage results at low primary system pressure.
PDS-32 Short-term SBO sequence involving a loss of high-pressure         4.OOE-09           0.06 injection due to two stuck-open safety relief valves or failure of HPCI and RCIC and one stuck-open safety relief valve.
Late injection and containment heat removal are available.
Core damage results at low primary system pressure. All accident-mitigating functions are recoverable when offsite power is restored.
However, containment venting is not available.
E.1-40
PDS-231 Short-term transients with loss of core cooling.
 
Core O.OOE+00 0.00 damage results at low primary system pressure.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report          Pilgrim LR Proceeding SAMA Analysis                                  50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.18 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
Late injection and containment venting are available, but containment heat removal is not available.
PDS                             Description Point Estimate           % of CDF PDS-33   Short-term large reactor vessel rupture. The resulting loss   4.OOE-09             0.06 of coolant is beyond the makeup capability of ECCS. Core damage occurs in the short term at low primary system pressure. Vessel injection and all forms of containment heat removal (RHR and containment venting) are available.
PDS-24 Similar to PDS-23, except that containment venting is not 4.98E-09 0.08 available.
I Cl E.1-39 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
Point PDS Description Estimate C of CDF PDS-25 Short-term transients with loss of core cooling. Core 2.57E-09 0.04 damage results at low primary system pressure.
No late injection, but containment heat removal and containment venting are available.
PDS-26 Similar to PDS-25, except that containment venting is not 1.24E-08 0.19 available.PDS-27 Short-term transients with loss of core cooling. Core 4.40E-11 0.001damage results at low primary system pressure.
Late injection and containment heat removal are not available.
However, containment venting is available PDS-28 Short-term transients with loss of core cooling. Core 1.10E-09 0.02 damage results at low primary system pressure.
Late injection, containment heat removal and containment venting are not available.
PDS-29 Long-term SBO involving loss of injection at high primary 1.41 E-07 2.21 system pressure from battery depletion.
All accident-mitigating functions are recoverable when AC power is restored.PDS-30 Short-term SBO sequence involving a loss of high-pressure O.OOE+00 0.00 injection at high primary system pressure from loss of all AC power and DC power or failure of SRVs. All accident-mitigating functions are recoverable when offsite power is restored.PDS-31 Long-term SBO sequence Involving a loss of high-pressure 2.60E-09 0.04 injection due to one stuck-open safety relief valve or long-term failure of HPCI and RCIC and subsequent failure to depressurize the primary system. Core damage results at low primary system pressure. All accident-rnitigating functions are recoverable when offsite power is restored.PDS-32 Short-term SBO sequence involving a loss of high-pressure 4.OOE-09 0.06 injection due to two stuck-open safety relief valves or failure of HPCI and RCIC and one stuck-open safety relief valve.Core damage results at low primary system pressure.
Allaccident-mitigating functions are recoverable when offsite power is restored.E.1-40 Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Table E.18 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
Point PDS Description Estimate % of CDFPDS-33 Short-term large reactor vessel rupture. The resulting loss 4.OOE-09 0.06 of coolant is beyond the makeup capability of ECCS. Coredamage occurs in the short term at low primary system pressure.
Vessel injection and all forms of containment heat removal (RHR and containment venting) are available.
The containment is not bypassed and AC power is available.
The containment is not bypassed and AC power is available.
PDS-34 Similar to PDS-33, except that containment heat removal O.OOE+00 0.00 from RHR fails.PDS-35 Short-term large reactor vessel rupture. The resulting loss O.OOE+00 0.00 of coolant is beyond the makeup capability of ECCS. Core damage occurs in the short term at low primary systempressure. Vessel injection is unavailable.
PDS-34   Similar to PDS-33, except that containment heat removal       O.OOE+00             0.00 from RHR fails.
However, all forms of containment heat removal (RHR and containment venting) are available.
PDS-35   Short-term large reactor vessel rupture. The resulting loss   O.OOE+00             0.00 of coolant is beyond the makeup capability of ECCS. Core damage occurs in the short term at low primary system pressure. Vessel injection is unavailable. However, all forms of containment heat removal (RHR and containment venting) are available. The containment is not bypassed PDS-36 and AC power is available.
The containment is not bypassed and AC power is available.
Similar to PDS-35, except that containment heat removal       0.OOE+00             0.00 CW from RHR fails.
PDS-36 Similar to PDS-35, except that containment heat removal 0.OOE+00 0.00from RHR fails.PDS-37 Short-term ATWS with failure of SRVs and SVs to open to- 1.95E-08 0.31 reduce primary system pressure.
PDS-37   Short-term ATWS with failure of SRVs and SVs to open to-     1.95E-08             0.31 reduce primary system pressure. The ensuing primary system over pressurization leads to a LOCA beyond core cooling capabilities. Late injection and containment heat removal are available.
The ensuing primary system over pressurization leads to a LOCA beyond corecooling capabilities.
PDS-38   Short-term ATWS that leads to early core damage at low       0.OOE+00           0.00 primary system pressure following successful reactivity control. Late injection is not available. However, containment heat removal is available.
Late injection and containment heat removal are available.PDS-38 Short-term ATWS that leads to early core damage at low 0.OOE+00 0.00 primary system pressure following successful reactivitycontrol. Late injection is not available.
PDS-39   Similar to PDS-38 except that containment heat removal       2.32E-09             0.04 from the RHR system is not available.
However, containment heat removal is available.
PDS-40   Long-term ATWS that leads to late core damage at low         0.OOE+00           0.00 primary system pressure following successful reactivity control. Late injection is available; containment heat removal from the RHR is not available. The containment is vented.
PDS-39 Similar to PDS-38 except that containment heat removal 2.32E-09 0.04from the RHR system is not available.
E.1-41 CJ
PDS-40 Long-term ATWS that leads to late core damage at low 0.OOE+00 0.00primary system pressure following successful reactivity control. Late injection is available; containment heatremoval from the RHR is not available. The containment is vented.CW.)E.1-41 C J Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
 
PDS Description Estinate %
Exhibit No. NRC000001 NRC - Applicant's Environmental Report             Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
of CDF PDS-41 Short-term ATWS that leads to early core damage at high 1.34E-11 <0.001primary system pressure following successful reactivity control. Late injection and containment heat removal are available.
PDS                             Description                               Estinate         % of CDF PDS-41   Short-term ATWS that leads to early core damage at high         1.34E-11           <0.001 primary system pressure following successful reactivity control. Late injection and containment heat removal are available.
PDS-42 Similar to PDS-41 except that containment heat removal 0.00E+00 0.00 from the RHR system Is not available.PDS-43 Long-term ATWS that leads to late core damage at high 0.OOE+00 0.00primary system pressure following successful reactivity control. Late injection is available; containment heatremoval from the RHR is not available. The containment is vented.PDS-44 Long-term ATWS that leads to late core damage at high 0.OOE+00 0.00 primary system pressure following successful reactivity control. Late injection is available.
PDS-42   Similar to PDS-41 except that containment heat removal           0.00E+00           0.00 from the RHR system Is not available.
However, containment heat removal from the RHR system and containment venting are not available.
PDS-43   Long-term ATWS that leads to late core damage at high           0.OOE+00           0.00 primary system pressure following successful reactivity control. Late injection is available; containment heat removal from the RHR is not available. The containment is vented.
PDS-45 Short-term ATWS that leads to containment failure and 3.39E-08 0.53 early core damage at high primary system pressure because of inadequate reactor water level following a loss of reactivity control. Late injection and containment venting are available.
PDS-44   Long-term ATWS that leads to late core damage at high           0.OOE+00           0.00 primary system pressure following successful reactivity control. Late injection is available. However, containment heat removal from the RHR system and containment venting are not available.
PDS-46 Short-term ATWS that leads to containment failure and 0.OOE+00 0.00 early core damage at high primary system pressure because of inadequate reactor water level following successful reactivity control. No late injection; however, containment venting Is available.
PDS-45   Short-term ATWS that leads to containment failure and           3.39E-08           0.53 early core damage at high primary system pressure because of inadequate reactor water level following a loss of reactivity control. Late injection and containment venting are available.
PDS-47 Unisolated LOCA outside containment with early core melt 3.22E-09 0.05 at high RPV pressure.PDS-48 Unisolated LOCA outside containment with early core melt 7.73E-10 0.01 at low RPV pressure.E.1-42 Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage The PDS designators listed in Table E.1-8 represent the core damage end state categories from the Level 1 analysis that are grouped together as entry conditions for the Level 2 analysis.
PDS-46 Short-term ATWS that leads to containment failure and             0.OOE+00           0.00 early core damage at high primary system pressure because of inadequate reactor water level following successful reactivity control. No late injection; however, containment venting Is available.
The Level 2 accident progression for each of the PDS is then evaluated using a single CET todetermine the appropriate release category for each Level 2 sequence.
PDS-47 Unisolated LOCA outside containment with early core melt         3.22E-09           0.05 at high RPV pressure.
Each end state associated with a Level 2 sequence is assigned to one of the release categories depicted in Table E.1-7. Note, however, that since not all the Level 2 sequences associated with each Level 1 core damage class 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.
PDS-48 Unisolated LOCA outside containment with early core melt         7.73E-10           0.01 at low RPV pressure.
Rather, the sum of the Level 2 end state frequencies assigned to each release category determines the overall frequency of that release category.
E.1-42
The CET described in the Level 2 model determines the release category frequency attributed to each Level 1 core damage PDS.E.1.2.2.6 Collapsed Accident Progression Bins Source Terms The source term analysis results in hundreds of source terms for internal initiators, making calculation with the MACCS2 consequence model cumbersome.
 
Therefore, the source termswere grouped into a much smaller number of source term groups defined in terms of similarproperties, with a frequency weighted mean source term for each group.The consequence analysis source terms groups are represented by collapsed accident progression bins (CAPB). The CAPB were generated by sorting the accident progression bins for each of the forty-eight PDS on attributes of the accident: the occurrence of core damage, the occurrence of vessel breach, primary system pressure at vessel breach, the location of 0 containment failure, the timing of containment failure, and the occurrence of core-concreteinteractions. Descriptions of the CAPB are presented in Table E.1-9.E.1-43 Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions CAPB Number Description CAPB-1 [CD, No VB, No CF, No CCI]Core damage (CD) occurs, but timely recovery of RPV injection prevents vessel breach (No VB). Therefore, containment integrity is not challenged (No CF) andcore-concrete interactions are precluded (No CCI).
Exhibit No. NRC000001 NRC - Applicant's Environmental Report          Pilgrim LR Proceeding SAMA Analysis                                  50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage The PDS designators listed in Table E.1-8 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 then evaluated using a single CET to determine the appropriate release category for each Level 2 sequence. Each end state associated with a Level 2 sequence is assigned to one of the release categories depicted in Table E.1-7. Note, however, that since not all the Level 2 sequences associated with each Level 1 core damage class 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. The CET described in the Level 2 model determines the release category frequency attributed to each Level 1 core damage PDS.
However, the potential exists for in-vessel release to the environment due to containment design leakage.CAPB-2 [CD, VB, No CF, No CCI]Core damage (CD) occurs followed by -vessel breach (VB). Containment does not'fail structurally and is not vented (No CF). Ex-vessel releases are recovered,precluding core-concrete interactions (No CCI). Although containment does not fail, vessel breach does occur, therefore the potential exists for in- and ex-vessel releases to the environment due to containment design leakage. RPV pressure isnot important because, even though high pressure induced severe accident phenomena (such as direct containment heating [DCH]) occurs, containment does not fail.CAPB-3 [CD, VB, No CF, CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment does notfail structurally and is not vented (No CF). However, ex-vessel releases are not recovered in time, and therefore core-concrete interactions occur (CCI). RPV pressure is not important because, even though high pressure induced severe accident phenomena (such as direct containment heating [DCH]) occurs, containment does not fail, nor is the vent limit reached.CAPB-4 [CD, VB, Early CF, WW, RPV pressure >200 psig at VB, No CCII Core damage (CD) occurs followed by 'vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the torus (WW), above the water level. RPV pressure Is greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena
E.1.2.2.6   Collapsed Accident Progression Bins Source Terms The source term analysis results in hundreds of source terms for internal initiators, making calculation with the MACCS2 consequence model cumbersome. Therefore, the source terms were grouped into a much smaller number of source term groups defined in terms of similar properties, with a frequency weighted mean source term for each group.
[DCH] are possible).
The consequence analysis source terms groups are represented by collapsed accident progression bins (CAPB). The CAPB were generated by sorting the accident progression bins for each of the forty-eight PDS on attributes of the accident: the occurrence of core damage, the occurrence of vessel breach, primary system pressure at vessel breach, the location of containment failure, the timing of containment failure, and the occurrence of core-concrete 0
There are no core concrete interactions (No CCI) due to the' presence of an overlying pool of water.II i i E.1-44 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage C!Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)
interactions. Descriptions of the CAPB are presented in Table E.1-9.
CAPB Description Number CAPB-5 [CD, VB, Early CF, WW, RPV pressure <200 psig at VB, No CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the torus (WW), above the water level. RPV pressure is less than 200 psig at time of vessel breach; precluding high pressure inducedsevere accident phenomena.
E.1-43
There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.CAPB-6 [CD, VB, Early CF, WW, RPV pressure >200 psig at VB, CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the torus (WW), above the water level. RPV pressure is greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena
 
[DCH] are possible). Following containment failure, core-concrete interactions occur (CCI).CAPB-7 [CD, VB, Early CF, WW, RPV pressure <200 psig at VB, CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the torus (WW), above the water level.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions CAPB Number                                           Description CAPB-1   [CD, No VB, No CF, No CCI]
RPV pressure is less than 200 psig at time of vessel breach; precluding high pressure inducedsevere accident phenomena. Following containment failure, core-concrete interactions occur (CCI).CAPB-8 [CD, VB, Early CF, DW, RPV pressure >200 psig at VB, No CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the drywell or below the torus water line (DW). RPVpressure is greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena
Core damage (CD) occurs, but timely recovery of RPV injection prevents vessel breach (No VB). Therefore, containment integrity is not challenged (No CF) and core-concrete interactions are precluded (No CCI). However, the potential exists for in-vessel release to the environment due to containment design leakage.
[DCH] are possible).
CAPB-2   [CD, VB, No CF, No CCI]
There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.C)o E.1-45 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)
Core damage (CD) occurs followed by -vessel breach (VB). Containment does not' fail structurally and is not vented (No CF). Ex-vessel releases are recovered, precluding core-concrete interactions (No CCI). Although containment does not fail, vessel breach does occur, therefore the potential exists for in- and ex-vessel releases to the environment due to containment design leakage. RPV pressure is not important because, even though high pressure induced severe accident phenomena (such as direct containment heating [DCH]) occurs, containment does not fail.
CAPB Description Number CAPB-9 [CD, VB, Early CF, DW, RPV pressure <200 psig at VB, No CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails eitherbefore core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the drywell or below the torus water line (DW). RPV pressure is less than 200 psig at time of vessel breach; precluding high pressureinduced severe accident phenomena. There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.CAPB-10 [CD, VB, Early CF, DW, RPV pressure >200 psig at VB, CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails eitherbefore core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the drywell or below the torus water line (DW). RPV pressure is greater than 200 psig at time of vessel breach (this implies that highpressure induced severe accident phenomena
CAPB-3   [CD, VB, No CF, CCI]
[OCH] are possible).
Core damage (CD) occurs followed by vessel breach (VB). Containment does not fail structurally and is not vented (No CF). However, ex-vessel releases are not recovered in time, and therefore core-concrete interactions occur (CCI). RPV pressure is not important because, even though high pressure induced severe accident phenomena (such as direct containment heating [DCH]) occurs, containment does not fail, nor is the vent limit reached.
Following containment failure, core-concrete interactions occur (CCI).CAPB-11 [CD, VB, Early CF, DW, RPV pressure <200 psig at VB, CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the drywell or below the torus water line (DW). RPV pressure is less than 200 psig at time of vessel breach; precluding high pressure Induced severe accident phenomena. Following containment failure, core-concrete interactions occur (CCI).CAPB-12 [CD, VB, Late CF, WW, No CCI]Core damage JCD) occurs followed by vessel breach (VB). Containment fails late due to loss of containment heat removal (Late CF). Containment failure occurs inthe torus (WW), above the water level.
CAPB-4   [CD, VB, Early CF, WW, RPV pressure >200 psig at VB, No CCII Core damage (CD) occurs followed by 'vessel breach (VB). Containment fails either                   II before core damage, during core damage, or at vessel breach (Early CF).                           i  i Containment failure occurs in the torus (WW), above the water level. RPV pressure Is greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena [DCH] are possible). There are no core concrete interactions (No CCI) due to the' presence of an overlying pool of water.
RPV pressure is not important because high-pressure severe accident phenomena (such as DCH) did not fail containment.
E.1-44
There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.E.1-46 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage (I Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)
 
CAPB Description Number CAPB-13 [CD, VB, Late CF, WW, CCIXCore damage (CD) occurs followed by vessel breach (VB). Containment fails late (late CF) due to core-concrete interactions (CCI) after vessel breach. Containment failure occurs in the torus (WW), above the water level. RPV pressure is notimportant because high-pressure severe accident phenomena (such as DCH) did not fail containment.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report            Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage C!
CAPB-14 [CD, VB, Late CF, DW, No CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails late due to loss of containment heat removal (Late CF). Containment failure occurs in the drywell or below the torus water level (DW). RPV pressure is not important because high-pressure severe accident phenomena did not fail containment.
Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)
There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.CAPB-1 5[CD, VB, Late CF, DW, CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails late (late CF) due to core-concrete interactions (CCI) after vessel breach. Containment failure occurs in the drywell or below the torus water level (DW). RPV pressure is not important because high-pressure severe accident phenomena did not fail containment.
CAPB                                           Description Number CAPB-5   [CD, VB, Early CF, WW, RPV pressure <200 psig at VB, No CCI]
CAPB-16 [CD, VB, BYPASS, RPV pressure >200 psig, No CCI]Small break interfacing system LOCA outside containment occurs. Core damage (CD) and subsequent vessel breach (VB) results at high RPV pressure with abypassed containment.
Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).
There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.CAPB-17 [CD, VB, BYPASS, RPV pressure <200 psig, No CCI]Large break interfacing system LOCA outside containment occurs.
Containment failure occurs in the torus (WW), above the water level. RPV pressure is less than 200 psig at time of vessel breach; precluding high pressure induced severe accident phenomena. There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.
Core damage (CD) and subsequent vessel breach (VB) results at low RPV pressure with a bypassed containment.
CAPB-6   [CD, VB, Early CF, WW, RPV pressure >200 psig at VB, CCI]
There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.E.1-47 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)
Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).
NuCber Description CAPB-18 [CD, VB, BYPASS, RPV pressure >200 psig, CCI]Small break interfacing system LOCA outside containment occurs. Core damage (CD) and subsequent vessel breach (VB) results at high RPV pressure with a bypassed containment.
Containment failure occurs in the torus (WW), above the water level. RPV pressure is greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena [DCH] are possible). Following containment failure, core-concrete interactions occur (CCI).
Following vessel breach, core-concrete interaction occurs (CCI).CAPB-19 [CD, VB, BYPASS, RPV pressure <200 psig, CCI]Large break interfacing system LOCA outside containment occurs.
CAPB-7   [CD, VB, Early CF, WW, RPV pressure <200 psig at VB, CCI]                                   C)o Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).
Core damage (CD) and subsequent vessel breach (VB) results at low RPV pressure with a bypassed containment. Following vessel breach, core-concrete interaction occurs (CCI).I l Based on the above binning methodology, the salient Level 2 results are summarized in Tables%mv E.1-10 and E.1-11 respectively. Table E.1-10 summarizes the results of the CET quantification.This table identifies the total annual release frequency for each Level 2 release category.Table E.1-11 provides the frequency, time, duration, energy, and elevation of release for each CAPB.E.1-48 Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal StageTable E.1-10 Summary of PNPS Containment Event Tree Quantification Release Category Release Frequency (Timing/Magnitude)
Containment failure occurs in the torus (WW), above the water level. RPV pressure is less than 200 psig at time of vessel breach; precluding high pressure induced severe accident phenomena. Following containment failure, core-concrete interactions occur (CCI).
(/RY)Late Low 4.53E-06Late Medium 1.56E-06 Late High O.OOE-00 Early Low 3.32E-08 Early Medium 6.48E-08 Early High 1.13E-07 No Containment Failure 1.11E-07 Nomenclature Timing L (Late) -Greater than 24 hours E (Early) -Less than 24 hours Magnitude JW)NCF LO MED Hi(Little to no release)(Low)(Medium)(High)-Less than 0.001% Csl-0.001 to 1% Cs1-1 to 10% Csl-Greater than 1 0% Csl E.1-49 ("V Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-11 Collapsed Accident Progression Bin (CAPB) Source Terms !-CAPB CAPB Frequency (Iyear)Warning Time (sec)Elevation (m)Release Release Start I Duration Release Energy (sec)(sec)(W)1 CAPB-1 9.51 E-08 3.98E+03 3.OOE+01 2.20E+04 9.OOE+03 2.61E+05 2 CAPB-2 1.27E-08 3.96E+03 3.OOE+01 2.20E+04 9.OOE+03 2.50E+05 3 CAPB-3 2.39E-09 3.96E+03 3.OOE+01 2.20E+04 9.OOE+03 2.50E+05 4 CAPB-4 3.29E-09 7.96E+03 3.OOE+01 1.83E+04 3.56E+03 1.IOE+07 5 CAPB-5 2.73E-09 1.31 E+04 3.OOE+01 2.53E+04 7.93E+03 8.34E+06 6 CAPB-6 7.95E-09 1.33E+04 3.OOE+01 2.56E+04 8.11E+03 8.23E+06 7 CAPB-7 7.93E-09 1.38E+04 3.OOE+01 2.61 E+04 8.46E+03 8.03E+06 8 CAPB-8 2.06E-08 9.18E+03 3.00E+01 2.OOE+04 4.59E+03 1.04E+07 9 CAPB-9 9.25E-09 9.21 E+03 3.OOE+01 2.44E+04 8.87E+03 4.18E+06 10 CAPB-10 8.53E-08 1.37E+04 3.OOE+01 2.60E+04 8.40E+03 8.06E+06 11 CAPB-11 4.35E-08 1.37E+04 3.OOE+01 2.60E+04 8.40E+03 8.06E+06 12 CAPB-12 1.70E-06 2.84E+04 3.OOE+01 4.64E+04 9.OOE+03 7.59E+06 13 CAPB-13 2.30E-09 9.14E+03 3.OOE+01 2.71E+04 9.OOE+03 1.80E+06 14 CAPB-14 2.26E-06 2.66E+04 3.OOE+01 4.46E+04 9.OOE+03 7.08E+06 15 CAPB-15 2.12E-06 2.81 E+04 3.OOE+01 4.62E+04 9.OOE+03 7.60E+06 16 CAPB-16 1.18E-09 3.96E+03 3.OOE+01 2.12E+04 9.OOE+03 2.50E+05 17 CAPB-17 6.91E-09 3.96E+03 3.OOE+01 2.14E+04 9.OOE+03 2.50E+05 18 CAPB-18 4.61E-10 3.96E+03 3.OOE+01 2.12E+04 9.OOE+03 2.50E+05 19 CAPB-19 2.43E-08 3.96E+03 3.OOE+01 2.18E+04 9.OOE+03 2.50E+05 E.1-50:
CAPB-8   [CD, VB, Early CF, DW, RPV pressure >200 psig at VB, No CCI]
Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-11 Collapsed Accident Progression Bin (CAPB) Source Terms (continued)
Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).
I Release Fractions NG T Cs Te Sr Ru La Ce Ba 1 1.99E-07 1.85E-07 1.85E-07 O.OOE+O0 1.24E-09 8.OOE-09 5.01E11 8.43E-11 1.70E-08 2 9.97E-05 4.81 E-05 4.66E-05 1.76E-07 3.97E-07 4.OOE-06 1.65E-08 5.15E-08 4.87E-06 3 9.97E-05 5.37E-05 4.97E-05 1.76E-06 5.80E-07 4.OOE-06 2.37E-08 1.57E-07 4,95E-06 4 1.OOE+00 4.90E-02 2.62E-02 4.18E-05 2.46E-05 3.66E-04 8.97E-07 3.04E-06 1.92E-04 5 9.85E-01 7.86E-02 3.68E-02 4.28E-05 4.1OE-05 3.66E-04 1.56E-06 6.79E-06 3.44E-04 6 1.OOE+00 4.02E-02 2.32E-02 1.48E-03 3.19E-04 3.66E-04 6.50E-06 7.17E-05 3.23E-047 9.76E-01 6.11 E-02 2.94E-02 1.26E-03 2,30E-04 3.66E-04 9.1 OE-06 1.06E-04 4.52E-04 8 1.OOE+00 2.98E-01 2.72E-01 3.07E-05 9.89E-04 2.23E-02 4.49E-05 6.57E-05 1.1 5E-02 9 5.97E-01 7.61 E-02 7.07E-02 1.41 E-05 9.72E-04 1.09E-02 3.69E-05 7.63E-05 1.02E-02 10 1.OOE+00 2.80E-01 2.49E-01 1.1 E-02 3.07E-03 1.81E-02 7.95E-05 5.81 E-04 1.03E-02 11 9.79E-01 1.73E-01 1.41 E-01 9.97E-03 3.13E-03 1.78E-02 1.22E-04 9.39E-04 1.72E-02 12 2.01 E-01 5.84E-05 4.37E-05 1.25E-07 2.36E-07 1.72E-06 8.04E-09 2.56E-08 2.99E-06 13 9.97E-01 7.99E-03 5.99E-03 1.76E-04 3.63E-05 3.66E-04 2.15E-06 1.41 E-05 4.52E-04 14 7.75E-01 2.88E-02 2.67E-02 2.47E-05 2.05E-04 2.13E-03 8.49E-06 2.27E-05 2.61 E-03 15 9.97E-01 2.76E-01 2.68E-41 1.27E-03 2.27E-03 2.25E-02 9.33E-05 3.OOE-04 2.74E-02 16 1.OOE+00 6.71 E-02 3.26E-02 4.06E-04 9.11 E-05 2.21 E-02 1.45E-06 1.65E-05 4.27E-05 17 9.72E-01 3.62E-01 3.37E-01 1.34E-03 2.37E-03 2.20E-02 9.90E-05 1.62E-04 8.57E-03 18 1.OOE+00 9.76E-02 6.25E-02 2.09E-02 4.67E-03 2.27E-02 7.45E-05 8.50E-04 2.12E-03 19 9.72E-01 4.03E-41 3.77E-01 6.87E-02 9.58E-03 2.26E-02 3.OOE-04 2.33E-03 1.20E-02 (-j E.1-51 Q-Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal StageE.1.2.2.7 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 failure occurs in the drywell or below the torus water line (DW). RPV pressure is greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena [DCH] are possible). There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.
* containment pressure and temperature versus time (time of containment failure isdetermined by comparing these values with the nominal containment capability);
E.1-45
* radionuclide release time and magnitude for a large number of radioisotopes; and* release fractions for twelve radionuclide species.
 
E.1.3 IPEEE Analysis E.1.3.1 Seismic Analysis PNPS performed a seismic PRA following the guidance of NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (lPEEE) for Severe Accident Vulnerabilities, June 1991. The seismic PRA model was performed in conjunction with the SQUG program in 1994 as part of the IPEEE submittal report [Reference E.1-6]. The seismic, high wind, and external flooding analyses determined that the plant is adequately designed to protect against the effects of these natural events.A number of plant improvements were identified in Table 2.4 of NUREG-1 742, Perspectives Gained from the IPEEE Program, Final Report, April 2002 [Reference E.1 -8]. Theseimprovements were implemented.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)
CAPB                                           Description Number CAPB-9     [CD, VB, Early CF, DW, RPV pressure <200 psig at VB, No CCI]
Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).
Containment failure occurs in the drywell or below the torus water line (DW). RPV pressure isless than 200 psig at time of vessel breach; precluding high pressure induced severe accident phenomena. There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.
CAPB-10   [CD, VB, Early CF, DW,RPV pressure >200 psig at VB, CCI]
Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).
Containment failure occurs in the drywell or below the torus water line (DW). RPV pressure is greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena [OCH] are possible). Following containment failure, core-concrete interactions occur (CCI).
CAPB-11   [CD, VB, Early CF, DW, RPV pressure <200 psig at VB, CCI]
Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).
Containment failure occurs in the drywell or below the torus water line (DW). RPV pressure is less than 200 psig at time of vessel breach; precluding high pressure Induced severe accident phenomena. Following containment failure, core-concrete interactions occur (CCI).
CAPB-12 [CD, VB, Late CF, WW, No CCI]
Core damage JCD) occurs followed by vessel breach (VB). Containment fails late due to loss of containment heat removal (Late CF). Containment failure occurs in the torus (WW), above the water level. RPV pressure is not important because high-pressure severe accident phenomena (such as DCH) did not fail containment.
There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.
E.1-46
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report            Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage (I
Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)
CAPB                                           Description Number CAPB-13   [CD, VB, Late CF, WW, CCIX Core damage (CD) occurs followed by vessel breach (VB). Containment fails late (late CF) due to core-concrete interactions (CCI) after vessel breach. Containment failure occurs in the torus (WW), above the water level. RPV pressure is not important because high-pressure severe accident phenomena (such as DCH) did not fail containment.
CAPB-14   [CD, VB, Late CF, DW, No CCI]
Core damage (CD) occurs followed by vessel breach (VB). Containment fails late due to loss of containment heat removal (Late CF). Containment failure occurs in the drywell or below the torus water level (DW). RPV pressure is not important because high-pressure severe accident phenomena did not fail containment. There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.
CAPB-1 5 [CD, VB, Late CF, DW, CCI]
Core damage (CD) occurs followed by vessel breach (VB). Containment fails late (late CF) due to core-concrete interactions (CCI) after vessel breach. Containment failure occurs in the drywell or below the torus water level (DW). RPV pressure is not important because high-pressure severe accident phenomena did not fail containment.
CAPB-16   [CD, VB, BYPASS, RPV pressure >200 psig, No CCI]
Small break interfacing system LOCA outside containment occurs. Core damage (CD) and subsequent vessel breach (VB) results at high RPV pressure with a bypassed containment. There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.
CAPB-17   [CD, VB, BYPASS, RPV pressure <200 psig, No CCI]
Large break interfacing system LOCA outside containment occurs. Core damage (CD) and subsequent vessel breach (VB) results at low RPV pressure with a bypassed containment. There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.
E.1-47
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)
NuCber                                               Description CAPB-18         [CD, VB, BYPASS, RPV pressure >200 psig, CCI]
Small break interfacing system LOCA outside containment occurs. Core damage (CD) and subsequent vessel breach (VB) results at high RPV pressure with a bypassed containment. Following vessel breach, core-concrete interaction occurs (CCI).
CAPB-19         [CD, VB, BYPASS, RPV pressure <200 psig, CCI]
Large break interfacing system LOCA outside containment occurs. Core damage (CD) and subsequent vessel breach (VB) results at low RPV pressure with a bypassed containment. Following vessel breach, core-concrete interaction occurs (CCI).
Il Based on the above binning methodology, the salient Level 2 results are summarized in Tables
%mv   E.1-10 and E.1-11 respectively. Table E.1-10 summarizes the results of the CET quantification.
This table identifies the total annual release frequency for each Level 2 release category.
Table E.1-11 provides the frequency, time, duration, energy, and elevation of release for each CAPB.
E.1-48
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                  Pilgrim LR Proceeding SAMA Analysis                                          50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage  (p-Table E.1-10 Summary of PNPS Containment Event Tree Quantification Release Category                           Release Frequency (Timing/Magnitude)                                   (/RY)
Late Low                                     4.53E-06 Late Medium                                     1.56E-06 Late High                                   O.OOE-00 Early Low                                     3.32E-08 Early Medium                                   6.48E-08 Early High                                   1.13E-07 No Containment Failure                               1.11E-07 Nomenclature Timing L   (Late) - Greater than 24 hours E   (Early) - Less than 24 hours JW)
Magnitude NCF     (Little to no release)     - Less than 0.001% Csl LO      (Low)                      - 0.001 to 1% Cs1 MED    (Medium)                    -1 to 10% Csl Hi      (High)                      - Greater than 10% Csl E.1-49
("V
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                        50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-11 Collapsed Accident Progression Bin (CAPB) Source Terms                 !
CAPB    Warning                        Release    Release        Release Elevation
    - CAPB   Frequency     Time                          Start  I Duration        Energy (m)
(Iyear)   (sec)                         (sec)     (sec)           (W) 1 CAPB-1     9.51 E-08 3.98E+03       3.OOE+01       2.20E+04   9.OOE+03       2.61E+05 2 CAPB-2     1.27E-08   3.96E+03       3.OOE+01       2.20E+04   9.OOE+03       2.50E+05 3 CAPB-3     2.39E-09   3.96E+03       3.OOE+01       2.20E+04   9.OOE+03       2.50E+05 4 CAPB-4     3.29E-09   7.96E+03       3.OOE+01       1.83E+04   3.56E+03       1.IOE+07 5 CAPB-5     2.73E-09   1.31 E+04     3.OOE+01       2.53E+04   7.93E+03       8.34E+06 6 CAPB-6     7.95E-09   1.33E+04       3.OOE+01       2.56E+04   8.11E+03       8.23E+06 7 CAPB-7     7.93E-09   1.38E+04       3.OOE+01       2.61 E+04   8.46E+03       8.03E+06 8 CAPB-8     2.06E-08   9.18E+03       3.00E+01       2.OOE+04   4.59E+03       1.04E+07 9 CAPB-9     9.25E-09   9.21 E+03     3.OOE+01       2.44E+04   8.87E+03       4.18E+06 10 CAPB-10   8.53E-08   1.37E+04       3.OOE+01       2.60E+04   8.40E+03       8.06E+06 11 CAPB-11   4.35E-08   1.37E+04       3.OOE+01       2.60E+04   8.40E+03       8.06E+06 12 CAPB-12   1.70E-06   2.84E+04       3.OOE+01       4.64E+04   9.OOE+03       7.59E+06 13 CAPB-13   2.30E-09   9.14E+03       3.OOE+01       2.71E+04   9.OOE+03       1.80E+06 14 CAPB-14   2.26E-06   2.66E+04       3.OOE+01       4.46E+04   9.OOE+03       7.08E+06 15 CAPB-15   2.12E-06   2.81 E+04     3.OOE+01       4.62E+04   9.OOE+03       7.60E+06 16 CAPB-16   1.18E-09   3.96E+03       3.OOE+01       2.12E+04   9.OOE+03       2.50E+05 17 CAPB-17   6.91E-09   3.96E+03       3.OOE+01       2.14E+04   9.OOE+03       2.50E+05 18 CAPB-18   4.61E-10   3.96E+03       3.OOE+01       2.12E+04   9.OOE+03       2.50E+05 19 CAPB-19   2.43E-08   3.96E+03       3.OOE+01       2.18E+04   9.OOE+03       2.50E+05 E.1-50:
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                        50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-11 Collapsed Accident Progression Bin (CAPB) Source Terms (continued)
I         Release Fractions NG               T   Cs         Te             Sr         Ru       La           Ce           Ba 1 1.99E-07 1.85E-07   1.85E-07 O.OOE+O0 1.24E-09 8.OOE-09             5.01E11     8.43E-11     1.70E-08 2 9.97E-05 4.81 E-05   4.66E-05   1.76E-07       3.97E-07     4.OOE-06 1.65E-08     5.15E-08 4.87E-06 3 9.97E-05 5.37E-05   4.97E-05   1.76E-06 5.80E-07 4.OOE-06 2.37E-08               1.57E-07     4,95E-06 4 1.OOE+00 4.90E-02 2.62E-02 4.18E-05             2.46E-05 3.66E-04     8.97E-07     3.04E-06     1.92E-04 5 9.85E-01 7.86E-02   3.68E-02 4.28E-05 4.1OE-05 3.66E-04             1.56E-06     6.79E-06     3.44E-04 6 1.OOE+00 4.02E-02   2.32E-02   1.48E-03 3.19E-04 3.66E-04           6.50E-06     7.17E-05     3.23E-04 7 9.76E-01 6.11 E-02 2.94E-02 1.26E-03 2,30E-04 3.66E-04             9.1 OE-06 1.06E-04 4.52E-04 8 1.OOE+00 2.98E-01   2.72E-01   3.07E-05       9.89E-04 2.23E-02 4.49E-05         6.57E-05     1.1 5E-02 9 5.97E-01 7.61 E-02 7.07E-02     1.41 E-05 9.72E-04         1.09E-02 3.69E-05     7.63E-05     1.02E-02 10 1.OOE+00 2.80E-01   2.49E-01   1.1 E-02       3.07E-03 1.81E-02     7.95E-05 5.81 E-04         1.03E-02 (-j 11 9.79E-01 1.73E-01   1.41 E-01   9.97E-03       3.13E-03     1.78E-02 1.22E-04     9.39E-04     1.72E-02 12 2.01 E-01 5.84E-05 4.37E-05     1.25E-07       2.36E-07 1.72E-06     8.04E-09 2.56E-08         2.99E-06 13 9.97E-01 7.99E-03   5.99E-03   1.76E-04       3.63E-05     3.66E-04 2.15E-06     1.41 E-05 4.52E-04 14 7.75E-01 2.88E-02   2.67E-02 2.47E-05         2.05E-04     2.13E-03 8.49E-06 2.27E-05         2.61 E-03 15 9.97E-01 2.76E-01   2.68E-41   1.27E-03       2.27E-03     2.25E-02 9.33E-05 3.OOE-04         2.74E-02 16 1.OOE+00 6.71 E-02 3.26E-02 4.06E-04           9.11 E-05 2.21 E-02   1.45E-06     1.65E-05     4.27E-05 17 9.72E-01 3.62E-01   3.37E-01   1.34E-03 2.37E-03 2.20E-02           9.90E-05     1.62E-04     8.57E-03 18 1.OOE+00 9.76E-02   6.25E-02 2.09E-02         4.67E-03 2.27E-02     7.45E-05 8.50E-04         2.12E-03 19 9.72E-01 4.03E-41   3.77E-01   6.87E-02       9.58E-03 2.26E-02     3.OOE-04     2.33E-03     1.20E-02 E.1-51 Q-
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report            Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.2.2.7     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 versus time (time of containment failure is determined by comparing these values with the nominal containment capability);
* radionuclide release time and magnitude for a large number of radioisotopes; and
* release fractions for twelve radionuclide species.
E.1.3   IPEEE Analysis E.1.3.1 Seismic Analysis PNPS performed a seismic PRA following the guidance of NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (lPEEE) for Severe Accident Vulnerabilities, June 1991. The seismic PRA model was performed in conjunction with the SQUG program in 1994 as part of the IPEEE submittal report [Reference E.1-6]. The seismic, high wind, and external flooding analyses determined that the plant is adequately designed to protect against the effects of these natural events.
A number of plant improvements were identified in Table 2.4 of NUREG-1 742, Perspectives Gained from the IPEEE Program, Final Report, April 2002 [Reference E.1 -8]. These improvements were implemented.
The seismic CDF in the IPEEE was conservatively estimated to be 5.82x10-5 per reactor-year.
The seismic CDF in the IPEEE was conservatively estimated to be 5.82x10-5 per reactor-year.
The seismic CDF has recently been re-evaluated to reflect the updated Gothic computer code room heat up calculations that predict no room cooling requirements for HPCI, RCIC, Core Spray, and RHR areas; to update random component failure probabilities; and to model replacement of certain relays with a seismically rugged model. The updated seismic CDF of 3.22x10-5 per reactor-year was used in estimation of the factor of 6 used to determine the upper bound estimated benefit described in Section 4.21.5.4.E.1.3.2 Fire Analysis The PNPS internal fire risk model was performed in 1994 as part of the IPEEE submittal report[Reference E.1-6]. The PNPS fire analysis was performed using the conservative EPRI's Fire Induced Vulnerability Evaluation (FIVE) methodology for qualitative and quantitative screening of fire areas and for fire analysis of areas that did not screen [Reference E.1 -71. The FIVE methodology is primarily a screening approach used to identify plant vulnerabilities due to fire initiating events.E.1-52 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 0Table E.1-12 presents the results of the PNPS IPEEE fire analysis. The values presented in Table E.1-12 are taken from NUREG-1742
The seismic CDF has recently been re-evaluated to reflect the updated Gothic computer code room heat up calculations that predict no room cooling requirements for HPCI, RCIC, Core Spray, and RHR areas; to update random component failure probabilities; and to model replacement of certain relays with a seismically rugged model. The updated seismic CDF of 3.22x10-5 per reactor-year was used in estimation of the factor of 6 used to determine the upper bound estimated benefit described in Section 4.21.5.4.
[Reference E.1-8]. These values are the same as the original IPEEE fire CDF results (2.20E-5 per reactor-year)
E.1.3.2 Fire Analysis The PNPS internal fire risk model was performed in 1994 as part of the IPEEE submittal report
[Reference E.1-6] after the response to NRC questions/issues regarding fire-modeling progression.
[Reference E.1-6]. The PNPS fire analysis was performed using the conservative EPRI's Fire Induced Vulnerability Evaluation (FIVE) methodology for qualitative and quantitative screening of fire areas and for fire analysis of areas that did not screen [Reference E.1 -71. The FIVE methodology is primarily a screening approach used to identify plant vulnerabilities due to fire initiating events.
A revised fire zone CDF of 1.91 E-5 per reactor-year, generated to reflect updated equipment failure probability andunavailability values was used in estimation of the factor of 6 used to determine the upper bound estimated benefit described in Section 4.21.5.4.
E.1-52
The significant fire scenarios involve fires occurring in the train B switchgear room, turbine building heater bay, vital motor generator set room, and train A switchgear room.Table E.1-12 PNPS Fire Updated Core Damage Frequency Results Fire New Compartment Description CDF/year Estimate Sub-Area CDF/year 1E Reactor Building West, El. 21 9.7E-07 8.25E-07 2B Turbine Building Heater Bay 2.1 E-06 2.74E-06 3A Train B RBCCW/TBCCW Pump and Heat 2.0E-06 1.31 E-06 Exchanger Room 4A Train A RBCCW[TBCCW Pump and Heat 9.8E-07 2.95E-07 Exchanger Room 6 Control Room 1.6E-06 8.90E-07 7 Cable Spreading Room 9.5E-07 7.85E-07 9 Vital Motor Generator Set Room 2.4E-06 2.38E-06 12 Train A Switchgear Room 3.1E-06 2.30E-06 13 Train B Switchgear Room 6.1E-06 6.85E-06 26 Main Transformer 1.5E-06 7.60E-07 2.2E-05 1.91 E-05 I E.1-53 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.3.3 Other External Hazards The PNPS IPEEE submittal
 
[Reference E.1-6], in addition to the internal fires and seismic events, examined a number of other external hazards:* high winds and tornadoes;
Exhibit No. NRC000001 NRC - Applicant's Environmental Report          Pilgrim LR Proceeding SAMA Analysis                                  50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 0
* external flooding; and* ice, hazardous chemical, transportation, and nearby facility incidents.
Table E.1-12 presents the results of the PNPS IPEEE fire analysis. The values presented in Table E.1-12 are taken from NUREG-1742 [Reference E.1-8]. These values are the same as the original IPEEE fire CDF results (2.20E-5 per reactor-year) [Reference E.1-6] after the response to NRC questions/issues regarding fire-modeling progression. A revised fire zone CDF of 1.91 E-5 per reactor-year, generated to reflect updated equipment failure probability and unavailability values was used in estimation of the factor of 6 used to determine the upper bound estimated benefit described in Section 4.21.5.4.
In consequence of the above external hazards evaluation, no plant modifications were required for PNPS.No risks to the plant occasioned by high winds and tornadoes, external floods, Ice, and hazardous chemical, transportation, and nearby facility incidents were identified that might lead to core damage with a predicted frequency in excess of 104 6/year. Therefore, these other external event hazards are not included in this attachment and are expected not to impact the conclusions of this SAMA evaluation.
The significant fire scenarios involve fires occurring in the train B switchgear room, turbine building heater bay, vital motor generator set room, and train A switchgear room.
E.1.4 PSA Model Peer Review and Difference between Current PSA Model and 1995 Update IPE E.1.4.1 PSA Model Peer Review The original IPE PSA model was peer reviewed on March 2000 using the BWROG PSA Peer Review Certification Implementation Guidelines. Facts and Observation sheets documented thecertification teams' insights and potential level of significance.
Table E.1-12 PNPS Fire Updated Core Damage Frequency Results Fire                                                                                     New Compartment                           Description                     CDF/year           Estimate Sub-Area                                                                                 CDF/year 1E         Reactor Building West, El. 21                         9.7E-07           8.25E-07 2B           Turbine Building Heater Bay                           2.1 E-06         2.74E-06 3A           Train B RBCCW/TBCCW Pump and Heat                     2.0E-06           1.31 E-06 I
As part of the update of the IPE PSA models, all major issues and observations from the BWROG Peer Review (i.e., Level A, B, C, and D observations) have been addressed and incorporated into the current IPE PSA model, April 2003 [Reference E.1-1].For the current IPE/PSA model update, individual work packages (event tree, fault tree, human reliability analysis (HRA), data, etc.) and internal flooding analysis were circulated to each PSA member for independent peer review. The accident sequence packages, system work packages, HRA, and internal flooding analyses were also assigned to the appropriate PNPSplant personnel for review. For example, event trees, system analyses, and fault tree models were forwarded to the applicable plant systems engineers and the HRA was assigned to individuals from the plant Operations Training department for review. Similarly, the accident sequence packages, system work packages, HRA report, containment performance analysis, fault tree and event tree models, and Level 2 models were peer reviewed by an outside consultant.
Exchanger Room 4A           Train A RBCCW[TBCCW Pump and Heat                     9.8E-07           2.95E-07 Exchanger Room 6           Control Room                                           1.6E-06           8.90E-07 7           Cable Spreading Room                                   9.5E-07           7.85E-07 9           Vital Motor Generator Set Room                         2.4E-06           2.38E-06 12         Train A Switchgear Room                               3.1E-06           2.30E-06 13         Train B Switchgear Room                               6.1E-06           6.85E-06 26           Main Transformer                                       1.5E-06           7.60E-07 2.2E-05           1.91 E-05 E.1-53
The Entergy license renewal project team and plant staff reviewed consequence and risk estimates for the SAMA analyses.E.1-54 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage The peer review process emphasized the role of plant staff, external consultants, and BWROG PSA certification in this recent model update. The peer reviews served to ensure the accuracy of both the assumptions made in the models and the results. The results of the peer review and resolutions are presented in Section 5 and Appendix P of the Pilgrim Nuclear Power Station Individual Plant Examination for Internal Events update report, April 2003 [Reference E.1-1].E.1.4.2 Major Differences between the Updated IPE PSA Model and 1995 Update IPE Model E.1.4.2.1 Core Damage -Comparison to the PNPS 1995 Update IPE Model The current PNPS IPE/PSA update model was completely revised in response to the BWROGPeer Review of March 2000 [Reference E.1-1]. The updated model is based upon all procedures and plant design as of September 30, 2001, and plant data as of December 31, 2001. The results yield a measurably lower CDF (point estimate CDF -6.41 E-6/reactor year) than the original IPE (point estimate CDF -5.85E-5/yr)
 
[Reference E.1-2] and 1995 PSA model update (point estimate CDF -2.84E-5/yr)
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.3.3 Other External Hazards The PNPS IPEEE submittal [Reference E.1-6], in addition to the internal fires and seismic events, examined a number of other external hazards:
[Reference E.1-31. (The 1995 update was performed to answer NRC questions following the IPE submittal.) The improved results are due to improved plant performance, replacement of switchyard -breakers, more realistic success criteria based on MAAP runs, and more sophisticated data handling. Major changes are summarized as follows.
* high winds and tornadoes;
A. Initiating Event The initiating event frequencies were updated to include current plant data and recent NRCpublication information.
* external flooding; and
For example, the LOOP frequency decreased significantly from the original IPE frequency of 0.475/yr to the current value of 0.067/yr [Reference E.1-1], whichreflects the decreased occurrence of LOOP events since 1990 and replacement of switchyard breakers.
* ice, hazardous chemical, transportation, and nearby facility incidents.
In addition, fault tree models were developed to calculate support system initiating event frequencies.
In consequence of the above external hazards evaluation, no plant modifications were required for PNPS.
B. Accident Sequence EvaluationEvent trees from the original IPE were completely revised. BWROG certification findings and observations were incorporated into the revised event trees. Major facts and observations include the following.
No risks to the plant occasioned by high winds and tornadoes, external floods, Ice, and hazardous chemical, transportation, and nearby facility incidents were identified that might lead to core damage with a predicted frequency in excess of 1046/year. Therefore, these other external event hazards are not included in this attachment and are expected not to impact the conclusions of this SAMA evaluation.
(1) LOOP Event Tree The LOOP event was completely revised to account for failure modes of HPCI/RCIC beyond 8 hours of operation; RPV depressurization on HCTL; and transfer to the SBOtree to address such items as premature battery depletion and AC recovery at 30 minutes and beyond.
E.1.4 PSA Model Peer Review and Difference between Current PSA Model and 1995 Update IPE E.1.4.1 PSA Model Peer Review The original IPE PSA model was peer reviewed on March 2000 using the BWROG PSA Peer Review Certification Implementation Guidelines. Facts and Observation sheets documented the certification teams' insights and potential level of significance. As part of the update of the IPE PSA models, all major issues and observations from the BWROG Peer Review (i.e., Level A, B, C, and D observations) have been addressed and incorporated into the current IPE PSA model, April 2003 [Reference E.1-1].
E.1-55 C>
For the current IPE/PSA model update, individual work packages (event tree, fault tree, human reliability analysis (HRA), data, etc.) and internal flooding analysis were circulated to each PSA member for independent peer review. The accident sequence packages, system work packages, HRA, and internal flooding analyses were also assigned to the appropriate PNPS plant personnel for review. For example, event trees, system analyses, and fault tree models were forwarded to the applicable plant systems engineers and the HRA was assigned to individuals from the plant Operations Training department for review. Similarly, the accident sequence packages, system work packages, HRA report, containment performance analysis, fault tree and event tree models, and Level 2 models were peer reviewed by an outside consultant.
Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage (2) SBO Event Tree Current update reflects GE load shed calculations and use of plant SBO procedures for DC load shedding.(3) Inadvertent Stuck Open Relief Valve (IORV)
The Entergy license renewal project team and plant staff reviewed consequence and risk estimates for the SAMA analyses.
Event Tree The IORV event tree was modified to include RPV depressurization with two SRVs given high-pressure injection failure.(4) LOCAs Event TreesThe update considers both HPCI and RCIC for small break LOCAs.Large and medium LOCAs and subsequent ATWS are modeled as core damage end states in the updated model. Small break LOCAs and ATWS are treated as similar to transient-induced ATWS.The vapor suppression system is considered during large LOCAs events.
E.1-54
(5) ATWS Event Tree The revised ATWS tree reflects the potential for MSIV closure on low RPV level.The revised ATWS model takes into consideration "inhibit ADS" and MSIV bypass issues. In addition, HRA values take into consideration ATWS accident progressions forRPV and containment conditions predicted by MAAR (6) Loss-of-Containment Heat Removal SequencesThe revised event trees model the potential impact from containment venting on low-pressure system operation.
 
For example, no credit is given for core spray and LPCI if containment venting is required.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report           Pilgrim LR Proceeding SAMA Analysis                                    50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage The peer review process emphasized the role of plant staff, external consultants, and BWROG PSA certification in this recent model update. The peer reviews served to ensure the accuracy of both the assumptions made in the models and the results. The results of the peer review and resolutions are presented in Section 5 and Appendix P of the Pilgrim Nuclear Power Station Individual Plant Examination for Internal Events update report, April 2003 [Reference E.1-1].
In addition, other containment related phenomena, such as high torus temperatures (HPCI) and high containment pressures (RCIC, SRVs)are reflected in the updated event trees.The update model only considers the DTV path for containment venting.
E.1.4.2 Major Differences between the Updated IPE PSA Model and 1995 Update IPE Model E.1.4.2.1   Core Damage     - Comparison to the PNPS 1995 Update IPE Model The current PNPS IPE/PSA update model was completely revised in response to the BWROG Peer Review of March 2000 [Reference E.1-1]. The updated model is based upon all procedures and plant design as of September 30, 2001, and plant data as of December 31, 2001. The results yield a measurably lower CDF (point estimate CDF - 6.41 E-6/reactor year) than the original IPE (point estimate CDF - 5.85E-5/yr) [Reference E.1-2] and 1995 PSA model update (point estimate CDF - 2.84E-5/yr) [Reference E.1-31. (The 1995 update was performed to answer NRC questions following the IPE submittal.) The improved results are due to improved plant performance, replacement of switchyard -breakers, more realistic success criteria based on MAAP runs, and more sophisticated data handling. Major changes are summarized as follows.
(7) ISLOCA Event Tree NSAC-154 [Reference E.1-10] and NUREG/CR-5124 [Reference E.1-11] were used to reassess the ISLOCA analysis.Success criteria for low-pressure injection during an ISLOCA are consistent with those used for small LOCAs.E.1-56 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal StageThe revised ISLOCA event tree credits use of condensate or fire water for large ISLOCA events provided that LPCI or core spray operation had previously occurred to provide initial RPV reflood.
A. Initiating Event The initiating event frequencies were updated to include current plant data and recent NRC publication information. For example, the LOOP frequency decreased significantly from the original IPE frequency of 0.475/yr to the current value of 0.067/yr [Reference E.1-1], which reflects the decreased occurrence of LOOP events since 1990 and replacement of switchyard breakers. In addition, fault tree models were developed to calculate support system initiating event frequencies.
B. Accident Sequence Evaluation Event trees from the original IPE were completely revised. BWROG certification findings and observations were incorporated into the revised event trees. Major facts and observations include the following.
(1) LOOP Event Tree The LOOP event was completely revised to account for failure modes of HPCI/RCIC beyond 8 hours of operation; RPV depressurization on HCTL; and transfer to the SBO tree to address such items as premature battery depletion and AC recovery at 30 minutes and beyond.
E.1-55                                                 C>
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report            Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage (2) SBO Event Tree Current update reflects GE load shed calculations and use of plant SBO procedures for DC load shedding.
(3) Inadvertent Stuck Open Relief Valve (IORV) Event Tree The IORV event tree was modified to include RPV depressurization with two SRVs given high-pressure injection failure.
(4) LOCAs Event Trees The update considers both HPCI and RCIC for small break LOCAs.
Large and medium LOCAs and subsequent ATWS are modeled as core damage end states in the updated model. Small break LOCAs and ATWS are treated as similar to transient-induced ATWS.
The vapor suppression system is considered during large LOCAs events.
(5) ATWS Event Tree The revised ATWS tree reflects the potential for MSIV closure on low RPV level.
The revised ATWS model takes into consideration "inhibit ADS" and MSIV bypass issues. In addition, HRA values take into consideration ATWS accident progressions for RPV and containment conditions predicted by MAAR (6) Loss-of-Containment Heat Removal Sequences The revised event trees model the potential impact from containment venting on low-pressure system operation. For example, no credit is given for core spray and LPCI if containment venting is required. In addition, other containment related phenomena, such as high torus temperatures (HPCI) and high containment pressures (RCIC, SRVs) are reflected in the updated event trees.
The update model only considers the DTV path for containment venting.
(7) ISLOCA Event Tree NSAC-154 [Reference E.1-10] and NUREG/CR-5124 [Reference E.1-11] were used to reassess the ISLOCA analysis.
Success criteria for low-pressure injection during an ISLOCA are consistent with those used for small LOCAs.
E.1-56
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report          Pilgrim LR Proceeding SAMA Analysis                                  50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage The revised ISLOCA event tree credits use of condensate or fire water for large ISLOCA events provided that LPCI or core spray operation had previously occurred to provide initial RPV reflood.
(8) Other Changes The revised event trees credit use of feedwater when appropriate.
(8) Other Changes The revised event trees credit use of feedwater when appropriate.
Control Rod Drive system flow into the RPV is credited for sequences that involve loss of containment heat removal and subsequent requirement to control containmentpressure with direct torus containment venting.
Control Rod Drive system flow into the RPV is credited for sequences that involve loss of containment heat removal and subsequent requirement to control containment pressure with direct torus containment venting.
Consistent success criteria were employed for RPV depressurization for transients, medium LOCAs, and small LOCAs.The revised PNPS IPE models are based on the BWROG EPGs/SAGs Revision 4 of the BWROG EPGs
Consistent success criteria were employed for RPV depressurization for transients, medium LOCAs, and small LOCAs.
[Reference E.1-1].Core damage definition has been revised to be consistent with the EPRI PSA Applications Guide [Reference E.1-12]. That is, core damage occurs when peak clad temperature exceeds 2200 0 F.HPCI and RCIC use is based on a 24-hour mission time.
The revised PNPS IPE models are based on the BWROG EPGs/SAGs Revision 4 of the BWROG EPGs [Reference E.1-1].
C C. Thermal -Hydraulic (T-H) Analysis T-H analysis has been completely revised and improved to better support the success criteria.The MAAP4 computer code [Reference E.1-4] was used to update and address the many issuesraised by the BWROG certification team, such as the following.
Core damage definition has been revised to be consistent with the EPRI PSA Applications Guide [Reference E.1-12]. That is, core damage occurs when peak clad temperature exceeds 22000F.
* A basis was provided for the timing and discharge pressure (flow) adequacy when usingthe fire water system for successful mitigation during transients and small LOCAs.* Success criteria for SORV are same as for non-SORV cases (2 SRVs are required for successful RPV depressurization).
HPCI and RCIC use is based on a 24-hour mission time.                                           C C. Thermal - Hydraulic (T-H) Analysis T-H analysis has been completely revised and improved to better support the success criteria.
* Consistent success criteria are used for RPV depressurization for transients, medium LOCAs, and small LOCAs.* Plant specific calculations were performed to identify the plant response for single ordouble recirculation pump trip failures.
The MAAP4 computer code [Reference E.1-4] was used to update and address the many issues raised by the BWROG certification team, such as the following.
* The appropriateness of the core damage definition used in the update was verified.E.1-57 Pilgrim Nuclear Power Station Applicant's Environmental Report Hi'" Operating License Renewal Stage In addition to the MAAP4 code, the GOTHIC code
* A basis was provided for the timing and discharge pressure (flow) adequacy when using the fire water system for successful mitigation during transients and small LOCAs.
[Reference E.1-13] was used to predict various room heatup rates for the reactor building, turbine building, switchgear room, and battery room.D. System Analysis System fault tree models from the original IPE were completely revised to reflect the as-built plant configuration.
* Success criteria for SORV are same as for non-SORV cases (2 SRVs are required for successful RPV depressurization).
MAAP analyses were clearly identified to support the success criteria of these Level 1 models. More detailed modeling for the logic interlock was included in the system models. A detailed fault tree for the RPS was developed based on NUREG/CR-5500
* Consistent success criteria are used for RPV depressurization for transients, medium LOCAs, and small LOCAs.
[Reference E. 1-9], which decreased the failure-to-scram probability from 3.OE-5/yr to 5.8E-6/yr.
* Plant specific calculations were performed to identify the plant response for single or double recirculation pump trip failures.
E Data Analysis Component failure data, both generic and plant-specific, were reviewed and updated with morerecent experience (the performance of risk significant systems HPCI and RCIC has greatly improved since the original IPE). Plant-specific data were adjusted for industry experience using Bayesian updates. Maintenance unavailability values were updated based on maintenance rule records from the system engineers.
* The appropriateness of the core damage definition used in the update was verified.
More recent common cause failure data and approach NUREG/CR-5497
E.1-57
[Reference E.1-14] were factored into this update. In particular, a more detailed and refined common-cause failure methodology (Alpha model) has been applied in this update. In addition, more common-cause equipment failure groups such as fans, dampers,/4gw transformers, DC power panels, and circuit breakers have been included in the analysis.F. HA A complete revision of the HRA was performed to identify, quantify, and document the pre-initiator and post-initiator human errors (including recoveries). The updated HRA was performedusing NUREG/CR-1278
 
[Reference E.1-15], also referred to as THERP. Screening values were only used for low-significance human errors. In addition, a detailed analysis was performed to treat dependencies between post-initiator errors.G Dependencv Analysis A complete revision of the internal flooding analysis was developed to systematically address spatial dependencies.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report            Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Hi'"                                                                             Operating License Renewal Stage In addition to the MAAP4 code, the GOTHIC code [Reference E.1-13] was used to predict various room heatup rates for the reactor building, turbine building, switchgear room, and battery room.
Dependency between pre-initiator human errors (such as miscalibration of instruments) was modeled. In addition, dependencies between multiple post-accident operator actions appearing in the same accident sequence were evaluated.  
D. System Analysis System fault tree models from the original IPE were completely revised to reflect the as-built plant configuration. MAAP analyses were clearly identified to support the success criteria of these Level 1 models. More detailed modeling for the logic interlock was included in the system models. A detailed fault tree for the RPS was developed based on NUREG/CR-5500 [Reference E.1-9], which decreased the failure-to-scram probability from 3.OE-5/yr to 5.8E-6/yr.
-Detailed component dependency tables were developed to address the support systems associated with the modeled systems and components.
E DataAnalysis Component failure data, both generic and plant-specific, were reviewed and updated with more recent experience (the performance of risk significant systems HPCI and RCIC has greatly improved since the original IPE). Plant-specific data were adjusted for industry experience using Bayesian updates. Maintenance unavailability values were updated based on maintenance rule records from the system engineers. More recent common cause failure data and approach NUREG/CR-5497 [Reference E.1-14] were factored into this update. In particular, a more detailed and refined common-cause failure methodology (Alpha model) has been applied in this update. In addition, more common-cause equipment failure groups such as fans, dampers,
E.1-58 Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage H. Structural ResponseThe ISLOCA frequency was revised.RPV overpressure and capability of the reactor building were included in the Level 2 assessment.
/4gw   transformers, DC power panels, and circuit breakers have been included in the analysis.
: 1. Quantification The truncation value was lowered to I.OE-11.Human Error Probability (HEP) dependencies and recovery actions in the cutsets were evaluated.
F. HA A complete revision of the HRA was performed to identify, quantify, and document the pre-initiator and post-initiator human errors (including recoveries). The updated HRA was performed using NUREG/CR-1278 [Reference E.1-15], also referred to as THERP. Screening values were only used for low-significance human errors. In addition, a detailed analysis was performed to treat dependencies between post-initiator errors.
ATWS contribution decreased due to lower probability of failure to scram based on NUREG/CR-5500 [Reference E.1-9].The HRA was completely revised to address a comment from the PSA Certification
G Dependencv Analysis A complete revision of the internal flooding analysis was developed to systematically address spatial dependencies.
[Reference E.1-16] that many of the HEPs were not realistic using the previous methodology.
Dependency between pre-initiator human errors (such as miscalibration of instruments) was modeled. In addition, dependencies between multiple post-accident operator actions appearing in the same accident sequence were evaluated. -
In many cases (e.g., failure to perform DTV), the previous HEPs were judged to be overly conservative.
Detailed component dependency tables were developed to address the support systems associated with the modeled systems and components.
J. Internal Flooding AnalysisThe internal flooding analysis from the original IPE was completely revised to include a detailed, systematic examination of the flood source and progression for each of the analyzed flooding scenarios.
E.1-58
In addition, the updated internal flooding analysis considers the effects of spray on equipment.
 
K. Uncertainty Analysis An uncertainty analysis was performed for this update.E.1.4.2.2 Containment Performance  
Exhibit No. NRC000001 NRC - Applicant's Environmental Report          Pilgrim LR Proceeding SAMA Analysis                                  50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage H. Structural Response The ISLOCA frequency was revised.
-Comparison to the Original PNPS IPE Model Containment performance analysis models were completely revised from the original IPE.Propagation of Level 1 cutsets to the Level 2 CET was developed.
RPV overpressure and capability of the reactor building were included in the Level 2 assessment.
A detailed LERF model was developed to ensure that LERF calculations are consistent with the PSA Applications Guide and NRC requirements for RG 1.174 [Reference E.1-17]. Other salient items incorporated are the following.
: 1. Quantification The truncation value was lowered to I.OE-11.
* CET fault models were revised to ensure that mitigating systems were not degraded in the Level I sequence.* CET fault tree models allowed credit for AC power recovery post core damage. This ensures that the models do not allow SBO core damage sequences to benefit from AC supported equipment in Level 2 without explicit consideration of AC power recovery.E.1-59 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage* Shell melt-through phenomena were considered where applicable.
Human Error Probability (HEP) dependencies and recovery actions in the cutsets were evaluated.
ATWS contribution decreased due to lower probability of failure to scram based on NUREG/CR-5500 [Reference E.1-9].
The HRA was completely revised to address a comment from the PSA Certification [Reference E.1-16] that many of the HEPs were not realistic using the previous methodology. In many cases (e.g., failure to perform DTV), the previous HEPs were judged to be overly conservative.
J. Internal Flooding Analysis The internal flooding analysis from the original IPE was completely revised to include a detailed, systematic examination of the flood source and progression for each of the analyzed flooding scenarios. In addition, the updated internal flooding analysis considers the effects of spray on equipment.
K. Uncertainty Analysis An uncertainty analysis was performed for this update.
E.1.4.2.2     Containment Performance     - Comparison to the Original PNPS IPE Model Containment performance analysis models were completely revised from the original IPE.
Propagation of Level 1 cutsets to the Level 2 CET was developed. A detailed LERF model was developed to ensure that LERF calculations are consistent with the PSA Applications Guide and NRC requirements for RG 1.174 [Reference E.1-17]. Other salient items incorporated are the following.
* CET fault models were revised to ensure that mitigating systems were not degraded in the Level I sequence.
* CET fault tree models allowed credit for AC power recovery post core damage. This ensures that the models do not allow SBO core damage sequences to benefit from AC supported equipment in Level 2 without explicit consideration of AC power recovery.
E.1-59
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report            Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage
* Shell melt-through phenomena were considered where applicable.
* Operator responses to key actions were reassessed to incorporate the probability for success given the containment conditions and Emergency Operating Procedure directions.
* Operator responses to key actions were reassessed to incorporate the probability for success given the containment conditions and Emergency Operating Procedure directions.
* Direct torus venting was considered post core damage.
* Direct torus venting was considered post core damage.
* PNPS-specific primary containment structural evaluation was included in the CET. This also included a structural evaluation of torus failure due to dynamic loading during ATWSscenarios, torus break below the water line, and bellows seal capability.
* PNPS-specific primary containment structural evaluation was included in the CET. This also included a structural evaluation of torus failure due to dynamic loading during ATWS scenarios, torus break below the water line, and bellows seal capability.
* A reactor building bypass fault tree model was developed to
* A reactor building bypass fault tree model was developed to assess the impact on the Level 2 analysis.
E.1.5 The MACCS2 Model - Level 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 the MACCS2 [Reference E.1-18] was created for PNPS. This model, which requires detailed site-specific meteorological, population, and economic data, estimates
A description of the analysis cases used in the evaluation follows.
A description of the analysis cases used in the evaluation follows.
Decay Heat Removal Capability  
Decay Heat Removal Capability - Torus Cooling This analysis case was used to evaluate the change in plant risk from installing an additional decay heat removal 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 the torus cooling mode of the RHR system to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $261,832. This analysis case was used to model the benefit of phase 11SAMAs 1 and 14.
-Torus Cooling This analysis case was used to evaluate the change in plant risk from installing an additional decay heat removal 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 the torus cooling mode of the RHR system to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $261,832. This analysis case was used to model the benefit of phase 11 SAMAs 1 and 14.Decay Heat Removal Capability  
Decay Heat Removal Capability - Drywell Sp=ra This analysis case was used to evaluate the change in plant risk from installing an additional decay heat removal system. Enhancements of decay heat removal capability decrease the E.2-4
-Drywell Sp=ra This analysis case was used to evaluate the change in plant risk from installing an additional decay heat removal system. Enhancements of decay heat removal capability decrease the E.2-4 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage probability of loss of containment heat removal. A bounding analysis was performed by setting the events for loss of the drywell spray mode of the RHR system to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately  
 
$264,219.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage probability of loss of containment heat removal. A bounding analysis was performed by setting the events for loss of the drywell spray mode of the RHR system to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $264,219. This analysis case was used to model the benefit of phase 11SAMA 9.
This analysis case was used to model the benefit of phase 11 SAMA 9.Filtered Vent This analysis case was used to evaluate the change in plant risk from installing a filtered containment vent to provide fission product scrubbing.
Filtered Vent This analysis case was used to evaluate the change in plant risk from installing a filtered containment vent to provide fission product scrubbing. A bounding analysis was performed by reducing the successful torus venting accident progression source terms by a factor of 2 to reflect the additional filtered capability. Reducing the releases from the vent path resulted in no benefit. This analysis case was used to model the benefit of phase 11SAMAs 2 and 19.
A bounding analysis was performed by reducing the successful torus venting accident progression source terms by a factor of 2 to reflect the additional filtered capability.
Containment Vent for ATWS Decay Heat Removal This analysis case was used to evaluate the change in plant risk from installing a containment vent to provide alternate decay heat removal capability during an ATWS event. A bounding analysis was performed by setting the ATWS sequences associated with containment bypass to zero in the level I PSA model, which resulted in an upper bound benefit of approximately
Reducing the releases from the vent path resulted in no benefit. This analysis case was used to model the benefit of phase 11 SAMAs 2 and 19.Containment Vent for ATWS Decay Heat Removal This analysis case was used to evaluate the change in plant risk from installing a containment vent to provide alternate decay heat removal capability during an ATWS event. A bounding analysis was performed by setting the ATWS sequences associated with containment bypass to zero in the level I PSA model, which resulted in an upper bound benefit of approximately
$61,701. This analysis case was used to model the benefit of phase 11SAMAs 3 and 47.
$61,701. This analysis case was used to model the benefit of phase 11 SAMAs 3 and 47.Molten Core Debris Removal This analysis case was used to estimate the change in plant risk from providing a molten core debris cooling mechanism.
Molten Core Debris Removal This analysis case was used to estimate the change in plant risk from providing a molten core debris cooling mechanism. A bounding analysis was performed by setting containment failure due to core-concrete interaction (not including liner failure) to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately $2,620,551. This analysis case was used to model the benefit of phase 11SAMAs 4, 5, 8, and 23.
A bounding analysis was performed by setting containment failure due to core-concrete interaction (not including liner failure) to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately  
Dryweff Head Flooding This analysis case was used to evaluate the change in plant risk from providing a modification to flood the drywell head such that if high drywell temperature occurred, the drywell head seal would not fail. A bounding analysis was performed by setting the probability of drywell head failure due to high temperature to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately $12,915. This analysis case was used to model the benefit of phase 11SAMAs 6,18, and 20.
$2,620,551.
Reactor Building Effectiveness This analysis case was used to evaluate the change in plant risk by ensuring the reactor building is available to provide effective fission product removal. Reactor building effectiveness was conservatively modeled by assuming reactor building availability for all accident sequences. This resulted in an upper bound benefit of approximately $64,577. This analysis case was used to model the benefit of phase II SAMAs 7, 13, and 21.
This analysis case was used to model the benefit of phase 11 SAMAs 4, 5, 8, and 23.Dryweff Head Flooding This analysis case was used to evaluate the change in plant risk from providing a modification to flood the drywell head such that if high drywell temperature occurred, the drywell head seal would not fail. A bounding analysis was performed by setting the probability of drywell head failure due to high temperature to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately  
E.2-5
$12,915. This analysis case was used to model the benefit of phase 11 SAMAs 6,18, and 20.Reactor Building Effectiveness This analysis case was used to evaluate the change in plant risk by ensuring the reactor building is available to provide effective fission product removal. Reactor building effectiveness was conservatively modeled by assuming reactor building availability for all accident sequences.
 
This resulted in an upper bound benefit of approximately  
Exhibit No. NRC000001 NRC - Applicant's Environmental Report          Pilgrim LR Proceeding SAMA Analysis                                  50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Strengthen Containment This analysis case was used to evaluate the change in plant risk from strengthening containment to reduce the probability of containment over-pressurization failure. A bounding analysis was performed by setting all energetic containment failure modes (DCH, steam explosions, late over-pressurization) to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately $1,233,428. This analysis case was used to model the benefit of phase 11SAMAs 10, 15, 16, and 24.
$64,577. This analysis case was used to model the benefit of phase II SAMAs 7, 13, and 21.E.2-5 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal StageStrengthen Containment This analysis case was used to evaluate the change in plant risk from strengthening containment to reduce the probability of containment over-pressurization failure.
Base Mat Melt-Through This analysis case was used to evaluate the change in plant risk from increasing the depth of the concrete base mat to ensure base mat melt-through does not occur. A bounding analysis was performed by setting containment failure due to base mat melt-through to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately $25,831. This analysis case was used to model the benefit of phase 11SAMA 11.
A bounding analysis wasperformed by setting all energetic containment failure modes (DCH, steam explosions, late over-pressurization) to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately  
Reactor Vessel Exterior Coolinm This analysis case was used to evaluate the change in plant risk from providing a method to perform ex-vessel cooling of the lower reactor vessel head. A bounding analysis was performed by modifying the probability of vessel failure by a factor of two to account for ex-vessel cooling in the level 2 PSA model, which resulted in an upper bound benefit of approximately $19,373. This analysis case was used to model the benefit of phase 11SAMA 12.
$1,233,428.
Vacuum Breakers This analysis case was used to evaluate the change in plant risk from improving the reliability of vacuum breakers to reseat following a successful opening and eliminate suppression pool scrubbing failures from the containment analysis. A bounding analysis was performed by setting the vacuum breaker failure probability to zero in the level 1 PSA model, which resulted in no benefit. This analysis case was used to model the benefit of phase 11SAMA 17.
This analysis case was used to model the benefit of phase 11 SAMAs 10, 15, 16, and 24.Base Mat Melt-Through This analysis case was used to evaluate the change in plant risk from increasing the depth of the concrete base mat to ensure base mat melt-through does not occur. A bounding analysis was performed by setting containment failure due to base mat melt-through to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately  
Flooding the Rubble Bed This analysis case was used to evaluate the change in plant risk from providing a source of water to the drywell floor to flood core debris. A bounding analysis was performed by substituting the probabilities of wet core concrete interactions for dry core concrete interactions in the level 2 PSA model, which resulted in an upper bound benefit of approximately $1,226,971. This analysis case was used to model the benefit of phase 11SAMA 22.
$25,831. This analysis case was used to model the benefit of phase 11 SAMA 11.Reactor Vessel Exterior CoolinmThis analysis case was used to evaluate the change in plant risk from providing a method to perform ex-vessel cooling of the lower reactor vessel head. A bounding analysis was performed by modifying the probability of vessel failure by a factor of two to account for ex-vessel cooling in the level 2 PSA model, which resulted in an upper bound benefit of approximately $19,373.
DC Power This analysis case was used to evaluate the change in plant risk from plant modifications that would increase the availability of Class 1E DC power (e.g., increasing battery capacity, using fuel cells, or extending SBO injection provisions). It was assumed that battery life could be extended E.2-6
This analysis case was used to model the benefit of phase 11 SAMA 12.Vacuum BreakersThis analysis case was used to evaluate the change in plant risk from improving the reliability of vacuum breakers to reseat following a successful opening and eliminate suppression pool scrubbing failures from the containment analysis.
 
A bounding analysis was performed by setting the vacuum breaker failure probability to zero in the level 1 PSA model, which resulted in no benefit. This analysis case was used to model the benefit of phase 11 SAMA 17.Flooding the Rubble BedThis analysis case was used to evaluate the change in plant risk from providing a source of water to the drywell floor to flood core debris.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage from 14 hours to 24 hours to simulate additional battery capacity. This enhancement would extend HPCI and RCIC operability and allow more credit for AC power recovery. A bounding analysis was performed by changing the time available to recover offsite power before HPCI and RCIC are lost from 14 hours to 24 hours during SBO scenarios in the level I PSA model. This resulted in an upper bound benefit of approximately $146,356. This analysis case was used to model the benefit of phase 11SAMAs 25, 26, 28, 33, and 35.
A bounding analysis was performed by substituting the probabilities of wet core concrete interactions for dry core concrete interactions in the level 2 PSA model, which resulted in an upper bound benefit of approximately  
Improve DC System This analysis case was used to evaluate the change in plant risk from improving injection capability by auto-transfer of AC bus control power to a standby DC power source upon loss of the normal DC source or from enhancing procedure to make use of DC bus cross-tie to improve DC power availability and reliability. A bounding analysis was performed by setting the DC buses D1 6 and D1 7 to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $118,568. This analysis case was used to model the benefit of phase 11SAMAs 27 and 34.
$1,226,971.
Altemate Pump Power Source This analysis case was used to evaluate the change in plant risk from adding a small, dedicated power source such as a dedicated diesel or gas turbine for the feedwater or condensate pumps so that they do not rely on offsite power. A bounding analysis was performed by setting failure of the SBO diesel generator to zero in level 1 PSA model, which resulted in an upper bound benefit of approximately $265,687. This analysis case was used to model the benefit of phase 11SAMA 29.
This analysis case was used to model the benefit of phase 11 SAMA 22.DC Power This analysis case was used to evaluate the change in plant risk from plant modifications that would increase the availability of Class 1 E DC power (e.g., increasing battery capacity, using fuel cells, or extending SBO injection provisions).
Improve AC Power System This analysis case was used to evaluate the change in plant risk from improving AC power system cross-tie capability to enhance the availability and reliability of the AC power system. A bounding analysis was performed by setting the loss of MCCs B17, B18, and B15 to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $473,410. This analysis case was used to model the benefit of phase 11SAMA 30.
It was assumed that battery life could be extended E.2-6 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage from 14 hours to 24 hours to simulate additional battery capacity.
Dedicated DC Power and Additional Batteries and Divisions This analysis case was used to evaluate the change in plant risk from plant modifications that would provide motive power to components (e.g., providing a dedicated DC power supply, additional batteries, or additional divisions). A bounding analysis was performed by setting the loss of DC bus D17 initiator, and one division of DC power, to zero in the level 1 PSA model, which resulted In an upper bound benefit of approximately $903,025. This analysis case was used to model the benefit of phase 11SAMAs 31 and 32.
This enhancement would extend HPCI and RCIC operability and allow more credit for AC power recovery.
E.2-7
A bounding analysis was performed by changing the time available to recover offsite power before HPCI and RCIC are lost from 14 hours to 24 hours during SBO scenarios in the level I PSA model. This resulted in an upper bound benefit of approximately  
 
$146,356.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report          Pilgrim LR Proceeding SAMA Analysis                                    50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Locate RHR Inside Containment This analysis case was used to evaluate the change in plant risk from moving the RHR system inside containment to prevent an RHR system ISLOCA event outside containment. A bounding analysis was performed by setting the RHR ISLOCA sequences to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $16,497. This analysis case was used to model the benefit of phase 11SAMA 36.
This analysis case was used to model the benefit of phase 11 SAMAs 25, 26, 28, 33, and 35.Improve DC System This analysis case was used to evaluate the change in plant risk from improving injection capability by auto-transfer of AC bus control power to a standby DC power source upon loss of the normal DC source or from enhancing procedure to make use of DC bus cross-tie to improve DC power availability and reliability. A bounding analysis was performed by setting the DC buses D1 6 and D1 7 to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $118,568. This analysis case was used to model the benefit of phase 11 SAMAs 27 and 34.Altemate Pump Power Source This analysis case was used to evaluate the change in plant risk from adding a small, dedicated power source such as a dedicated diesel or gas turbine for the feedwater or condensate pumps so that they do not rely on offsite power. A bounding analysis was performed by setting failure of the SBO diesel generator to zero in level 1 PSA model, which resulted in an upper bound benefit of approximately  
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. A bounding analysis was performed by setting the ISLOCA initiator to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $24,148. This analysis case was used to model the benefit of phase 11SAMA 37.
$265,687.
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 setting the containment bypass failure due to MSIV leakage to zero in the level 2 PSA model, which resulted in no benefit. This analysis case was used to model the benefit of phase 11SAMA 38.
This analysis case was used to model the benefit of phase 11 SAMA 29.Improve AC Power SystemThis analysis case was used to evaluate the change in plant risk from improving AC power system cross-tie capability to enhance the availability and reliability of the AC power system. A bounding analysis was performed by setting the loss of MCCs B17, B18, and B15 to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $473,410.
Diesel to CST Makeup Pumps This analysis case was used to evaluate the change in plant risk from installing an independent diesel for the CST makeup pumps to allow continued operation of the high pressure injection system during an SBO event. As currently modeled, if CST water level is low, swapping HPCI/
This analysis case was used to model the benefit of phase 11 SAMA 30.Dedicated DC Power and Additional Batteries and Divisions This analysis case was used to evaluate the change in plant risk from plant modifications thatwould provide motive power to components (e.g., providing a dedicated DC power supply, additional batteries, or additional divisions).
RCIC suction from the CST to the torus allows continued HPCI and RCIC injection. Therefore, a bounding analysis was performed by setting the failure to switchover from CST to torus to zero in the level 1 PSA model, which resulted in no benefit. This analysis case was used to model the benefit of phase 11SAMA 39.
A bounding analysis was performed by setting the loss of DC bus D17 initiator, and one division of DC power, to zero in the level 1 PSA model, which resulted In an upper bound benefit of approximately $903,025.
High Pressure Injection System This analysis case was used to evaluate the change in plant risk from plant modifications that would increase the availability of high pressure injection (e.g., installing an independent AC powered high pressure injection system, passive high pressure injection system, or an additional high pressure injection system). A bounding analysis was performed by setting the CDF contribution due to unavailability of the HPCI system to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $110,212. This analysis case was used to model the benefit of phase II SAMAs 40, 41, 42, 44, and 45.
This analysis case was used to model the benefit of phase 11 SAMAs 31 and 32.E.2-7 Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Locate RHR Inside ContainmentThis analysis case was used to evaluate the change in plant risk from moving the RHR systeminside containment to prevent an RHR system ISLOCA event outside containment.
E.2-8
A bounding analysis was performed by setting the RHR ISLOCA sequences to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately  
 
$16,497. This analysis case was used to model the benefit of phase 11 SAMA 36.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. A bounding analysis was performed by setting the ISLOCA initiator to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $24,148. This analysis case was used to model the benefit of phase 11 SAMA 37.MSIV Design This analysis case was used to evaluate the change in plant risk from improving MSIV design todecrease the likelihood of containment bypass scenarios.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report             Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Improve the Reliability of High Pressure Injection System This analysis case was used to evaluate the change in plant risk from plant modifications that would increase the reliability of the high pressure injection system. A bounding analysis was performed by reducing the HPCI system failure probability by a factor of three in the level 1 PSA model, which resulted in an upper bound benefit of approximately $76,025. This analysis case was used to model the benefit of phase 11SAMA 43.
A bounding analysis was performed by setting the containment bypass failure due to MSIV leakage to zero in the level 2 PSA model, which resulted in no benefit. This analysis case was used to model the benefit of phase 11 SAMA 38.Diesel to CST Makeup Pumps This analysis case was used to evaluate the change in plant risk from installing an independent diesel for the CST makeup pumps to allow continued operation of the high pressure injection system during an SBO event. As currently modeled, if CST water level is low, swapping HPCI/RCIC suction from the CST to the torus allows continued HPCI and RCIC injection. Therefore, a bounding analysis was performed by setting the failure to switchover from CST to torus to zero in the level 1 PSA model, which resulted in no benefit. This analysis case was used to model the benefit of phase 11 SAMA 39.High Pressure Injection SystemThis analysis case was used to evaluate the change in plant risk from plant modifications that would increase the availability of high pressure injection (e.g., installing an independent AC powered high pressure injection system, passive high pressure injection system, or an additional high pressure injection system). A bounding analysis was performed by setting the CDFcontribution due to unavailability of the HPCI system to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $110,212. This analysis case was used to model the benefit of phase II SAMAs 40, 41, 42, 44, and 45.
SRVs Reseat This analysis case was used to evaluate the change in plant risk from improving the reliability of SRVs reseating. A bounding analysis was performed by setting the stuck open SRVs initiator to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately
E.2-8 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Improve the Reliability of High Pressure Injection System This analysis case was used to evaluate the change in plant risk from plant modifications that would increase the reliability of the high pressure injection system. A bounding analysis was performed by reducing the HPCI system failure probability by a factor of three in the level 1 PSA model, which resulted in an upper bound benefit of approximately  
$63,599. This analysis case was used to model the benefit of phase 11SAMA 46.
$76,025. This analysis case was used to model the benefit of phase 11 SAMA 43.SRVs Reseat This analysis case was used to evaluate the change in plant risk from improving the reliability of SRVs reseating.
Diversity of Explosive Valves This analysis case was used to evaluate the change in plant risk from providing an alternate means of opening a pathway to the RPV for SLC system injection, thereby improving success probability for reactor shutdown. A bounding analysis was performed by setting common cause failure of SLC explosive valves to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $12,915. This analysis case was used to model the benefit of phase II SAMA48.
A bounding analysis was performed by setting the stuck open SRVs initiator to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately
Reliability of SRVs This analysis case was used to evaluate the change in plant risk from installing additional signals to automatically open the SRVs. This improvement would reduce the likelihood of SRVs failing to open, thereby reducing the consequences of medium LOCAs. A bounding analysis was performed by setting the probability of SRVs failing to open when required by reactor pressure vessel overpressure conditions to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $31,799. This analysis case was used to model the benefit of phase 11SAMA 49.
$63,599. This analysis case was used to model the benefit of phase 11 SAMA 46.Diversity of Explosive Valves This analysis case was used to evaluate the change in plant risk from providing an alternate means of opening a pathway to the RPV for SLC system injection, thereby improving success probability for reactor shutdown.
Improve SRV Design This analysis case was used to evaluate the change in plant risk from improving the SRV design to increase the reliability of opening, thus increasing the likelihood that accident sequences could be mitigated using low pressure injection systems. A bounding analysis was performed by setting the probability of SRVs failing to open during RPV depressurization to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $194,378. This analysis case was used to model the benefit of phase 11SAMA 50.
A bounding analysis was performed by setting common cause failure of SLC explosive valves to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately  
E.2-9
$12,915. This analysis case was used to model the benefit of phase II SAMA48.Reliability of SRVs This analysis case was used to evaluate the change in plant risk from installing additional signalsto automatically open the SRVs. This improvement would reduce the likelihood of SRVs failing to open, thereby reducing the consequences of medium LOCAs. A bounding analysis was performed by setting the probability of SRVs failing to open when required by reactor pressure vessel overpressure conditions to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately  
 
$31,799. This analysis case was used to model the benefit of phase 11 SAMA 49.Improve SRV DesignThis analysis case was used to evaluate the change in plant risk from improving the SRV design to increase the reliability of opening, thus increasing the likelihood that accident sequences could be mitigated using low pressure injection systems.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report          Pilgrim LR Proceeding SAMA Analysis                                  50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Self-Cooled ECCS Pump Seals This analysis case was used to evaluate the change in plant risk from providing self-cooled ECCS pump seals to eliminate dependence on the component cooling water system. A bounding analysis was performed by setting the CDF contribution from sequences involving RHR pump failures to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $29,412. This analysis case was used to model the benefit of phase 11SAMA 51.
A bounding analysis was performed by setting the probability of SRVs failing to open during RPV depressurization to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately  
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 break LOCA initiator to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $14,109. This analysis case was used to model the benefit of phase 11SAMA 52.
$194,378.
Controlled Containment Venting This analysis case was used to evaluate the change in plant risk from changing the design of the containment vent valves and procedure to establish a narrow pressure control band. This would prevent rapid containment depressurization when venting, thus avoiding adverse impact on the ability of the low pressure ECCS injection systems to take suction from the torus. A bounding analysis was performed by reducing the probability of the operator failing to recognize the need to vent the torus by a factor of three in the level 1 PSA model, which resulted in an upper bound benefit of approximately $137,237. This analysis case was used to model the benefit of phase 11 SAMA 53.
This analysis case was used to model the benefit of phase 11 SAMA 50.E.2-9 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Self-Cooled ECCS Pump SealsThis analysis case was used to evaluate the change in plant risk from providing self-cooled ECCS pump seals to eliminate dependence on the component cooling water system. A bounding analysis was performed by setting the CDF contribution from sequences involving RHR pump failures to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately  
ECCS Low Pressure Interlock This analysis case was used to evaluate the change in plant risk from installing a bypass switch to allow operator to bypass the ECCS low pressure interlock circuitry that inhibits opening of the RHR low pressure injection and core spray injection valves following sensor or logic failure. A bounding analysis was performed by setting the CDF contribution due to sensor failure, low pressure permissive logic failure, and miscalibration to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately $21,761. This analysis case was used to model the benefit of phase 11SAMA 54.
$29,412. This analysis case was used to model the benefit of phase 11 SAMA 51.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 break LOCA initiator to zero in the level 1 PSA model, which resulted in an upper bound benefit ofapproximately $14,109. This analysis case was used to model the benefit of phase 11 SAMA 52.Controlled Containment VentingThis analysis case was used to evaluate the change in plant risk from changing the design of the containment vent valves and procedure to establish a narrow pressure control band. This would prevent rapid containment depressurization when venting, thus avoiding adverse impact on the ability of the low pressure ECCS injection systems to take suction from the torus. A bounding analysis was performed by reducing the probability of the operator failing to recognize the need to vent the torus by a factor of three in the level 1 PSA model, which resulted in an upper bound benefit of approximately $137,237. This analysis case was used to model the benefit of phase 11 SAMA 53.ECCS Low Pressure Interlock This analysis case was used to evaluate the change in plant risk from installing a bypass switch to allow operator to bypass the ECCS low pressure interlock circuitry that inhibits opening of the RHR low pressure injection and core spray injection valves following sensor or logic failure. A bounding analysis was performed by setting the CDF contribution due to sensor failure, lowpressure permissive logic failure, and miscalibration to zero in the level 1 PSA model.
Improve the Reliability of SSW and RBCCW Pumps This analysis case was used to evaluate the change in plant risk from providing a separate pump train to eliminate common cause failure of SSW and RBCCW pumps. A bounding analysis was performed by setting the CDF contribution due to common cause failures of SSW and RBCCW pumps to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately $356,310. This analysis case was used to model the benefit of phase 11SAMA 55.
This resulted in an upper bound benefit of approximately  
E.2-10
$21,761. This analysis case was used to model the benefit of phase 11 SAMA 54.Improve the Reliability of SSW and RBCCW PumpsThis analysis case was used to evaluate the change in plant risk from providing a separate pump train to eliminate common cause failure of SSW and RBCCW pumps. A bounding analysis was performed by setting the CDF contribution due to common cause failures of SSW and RBCCW pumps to zero in the level 1 PSA model. This resulted in an upper bound benefit ofapproximately $356,310.
* Exhibit No. NRC000001 NRC - Applicant's Environmental Report              Pilgrim LR Proceeding SAMA Analysis                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Redundant DC Power Supplies to DTV Valves This analysis case was used to evaluate the change in plant risk from installing additional fuses to two DTV valve control circuits to enable the DTV function. A bounding analysis was performed by setting the CDF contribution due to DC power supply failures to DTV valves AO-5042B and AO-5025 to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately $220,639. -This analysis case was used to model the benefit of phase 11SAMA 56.
This analysis case was used to model the benefit of phase 11 SAMA 55.E.2-10
Proceduralize the Use of Diesel Fire Pump Hydroturbine This analysis case was used to evaluate the change in plant risk from revising the procedure to allow use of hydroturbine if EDG X-107A or diesel driven fire water pump P-140 is unavailable. A bounding analysis was performed by setting the CDF contribution from the sequences involving a LOOP and failure of either EDG A or fuel oil transfer oil pump (P-141) to zero in the level I PSA model. This resulted in an upper bound benefit of approximately $175,279. This analysis case was used to model the benefit of phase 11SAMA 57.
* Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Redundant DC Power Supplies to DTV Valves This analysis case was used to evaluate the change in plant risk from installing additional fuses to two DTV valve control circuits to enable the DTV function.
Proceduralize Alignment of Bus B3 to Feed Bus BI Loads or Bus B4 to Bus B2 This analysis case was used to evaluate the change in plant risk from providing a procedure to direct the operator to restore 480V MCCs B15 and B17 loads upon loss of 4.16kV bus A5 provided that 4.16kV bus A3 is available. The same is true for restoring 480V MCCs B14 and B18 loads upon loss of 4.16kV bus A6 provided that 4.16kV bus A4 is available. A bounding analysis was performed by setting the CDF contribution from the sequences involving a loss of the 4.16 kV bus A5 to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately $190,797. This analysis case was used to model the benefit of phase 11SAMA 58.
A bounding analysis was performedby setting the CDF contribution due to DC power supply failures to DTV valves AO-5042B and AO-5025 to zero in the level 1 PSA model.
Redundant Path from Fire Water Pump Discharge to LPCI Loops A and B Cross-tie This analysis case was used to evaluate the change in plant risk from installing a redundant path from fire protection water pump discharge to LPCI loops A and B cross-tie. A bounding analysis was performed by setting the CDF contribution from the sequences involving fire water into LPCI loops A and B cross-tie failure to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately $929,797. This analysis case was used to model the benefit of phase 11 SAMA 59.
This resulted in an upper bound benefit of approximately  
E.2.4 Sensitivity Analyses Two sensitivity analyses were conducted to gauge the impact of assumptions upon the analysis.
$220,639.  
The benefits estimated for each of these sensitivities are presented in Table E.2-2.
-This analysis case was used to model the benefit of phase 11 SAMA 56.Proceduralize the Use of Diesel Fire Pump Hydroturbine This analysis case was used to evaluate the change in plant risk from revising the procedure to allow use of hydroturbine if EDG X-107A or diesel driven fire water pump P-140 is unavailable.
A description of each sensitivity case follows.
A bounding analysis was performed by setting the CDF contribution from the sequences involving a LOOP and failure of either EDG A or fuel oil transfer oil pump (P-141) to zero in the level I PSA model. This resulted in an upper bound benefit of approximately $175,279. This analysis case was used to model the benefit of phase 11 SAMA 57.Proceduralize Alignment of Bus B3 to Feed Bus BI Loads or Bus B4 to Bus B2This analysis case was used to evaluate the change in plant risk from providing a procedure to direct the operator to restore 480V MCCs B15 and B17 loads upon loss of 4.16kV bus A5 provided that 4.16kV bus A3 is available.
E.2-11
The same is true for restoring 480V MCCs B14 and B18 loads upon loss of 4.16kV bus A6 provided that 4.16kV bus A4 is available.
 
A bounding analysis was performed by setting the CDF contribution from the sequences involving a loss of the 4.16 kV bus A5 to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately  
Exhibit No. NRC000001 NRC - Applicant's Environmental Report          Pilgrim LR Proceeding SAMA Analysis                                    50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's 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 27-year period for remaining plant life (i.e. seven years on the original plant license plus the 20-year license renewal period). The 20-year license renewal period was used in the base case. The resultant monetary equivalent was calculated using 27 years remaining until end of facility life to investigate the impact on each analysis case. Changing this assumption does not cause any additional SAMAs to be cost-beneficial.
$190,797.
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 any additional SAMAs to be cost-beneficial.
This analysis case was used to model the benefit of phase 11 SAMA 58.Redundant Path from Fire Water Pump Discharge to LPCI Loops A and B Cross-tie This analysis case was used to evaluate the change in plant risk from installing a redundant path from fire protection water pump discharge to LPCI loops A and B cross-tie.
E.2-12
A bounding analysis was performed by setting the CDF contribution from the sequences involving fire water into LPCI loops A and B cross-tie failure to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately  
 
$929,797.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report               Pilgrim LR Proceeding SAMA Analysis                                        50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.2.5 References E.2-1 Appendix D-Attachment F,Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Edwin I. Hatch Nuclear Power Plant Units 1 and 2, March 2000.
This analysis case was used to model the benefit of phase 11 SAMA 59.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.2-2.A description of each sensitivity case follows.E.2-11 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Sensitivity Case 1: Years Remaining Until End of Plant LifeThe purpose of this sensitivity case was to investigate the sensitivity of assuming a 27-year period for remaining plant life (i.e. seven years on the original plant license plus the 20-year license renewal period). The 20-year license renewal period was used in the base case. The resultant monetary equivalent was calculated using 27 years remaining until end of facility life to investigate the impact on each analysis case. Changing this assumption does not cause any additional SAMAs to be cost-beneficial.
E.2-2   U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Calvert Cliffs Nuclear Power Plant, Supplement 1, February 1999.
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 relativeto 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 any additional SAMAs to be cost-beneficial.
E.2-3 General Electric Nuclear Energy, Technical Support Document for the ABWR, 25A5680, Revision 1, January 18,1995.
E.2-12 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.2.5 References E.2-1 Appendix D-Attachment F, Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Edwin I. Hatch Nuclear Power Plant Units 1 and 2, March 2000.E.2-2 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Calvert Cliffs Nuclear Power Plant, Supplement 1, February 1999.E.2-3 General Electric Nuclear Energy, Technical Support Document for the ABWR, 25A5680, Revision 1, January 18,1995.E.24 Appendix E- Environmental Report, Appendix G, Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Peach Bottom Nuclear Power Plant Units 2 and 3, July, 2001.E.2-5 Appendix F, Severe Accident Mitigation Alternatives Analysis Submittal Related to Licensing Renewal for the Quad Cities Nuclear Power Plant Units 1 and 2, January 2003.E.2-6 Appendix F, Severe Accident Mitigation Alternatives Analysis Submittal Related to Licensing Renewal for the Dresden Nuclear Power Plant Units 2 and 3, January 2003.E.2-7 Appendix E-Attachment E, Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Arkansas Nuclear One -Unit 2, October 2003.E.2-8 Cost Estimate for Severe Accident Mitigation Design Alternatives, Limerick Generating Station for Philadelphia Electric Company, Bechtel Power Corporation, June 22, 1989.E.2-9 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Volume 1, 5.35, Listing of SAMDAs considered for the Limerick Generating Station, May 1996.E.2-10 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Volume 1, 5.36, Listing of SAMDAs considered for the Comanche Peak Steam Electric Station, May 1996.E.2-11 Museler, W. J., (Tennessee Valley Authority) to NRC Document Control Desk, "Watts Bar Nuclear Plant (WBN) Units I and 2 -Severe Accident Mitigation Design Alternatives (SAMDAs)," letter dated October 7, 1994.E.2-12 Nunn, D. E., (TVA) to NRC Document Control Desk, "Watts Bar Nuclear Plant (WBN)Units I and 2 -Severe Accident Mitigation Design Alternatives (SAMDA) -Response to Request for Additional Information (RAI) -(TAC Nos. M77222 and M77223)," letter dated October 7, 1994.E.2-13 Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage E.2-13 Liparulo, N. J., (Westinghouse Electric Corporation) to NRC Document Control Desk,"Submittal of Material Pertinent to the AP600 Design Certification Review," letter dated December 15,1992.E.2-14 U.S. Nuclear Regulatory Commission, NUREG-0498, Final Environmental Statement related to the operation of Watts Bar Nuclear Plant, Units 1 and 2, Supplement No. 1, April 1995.E.2-15 U.S. Nuclear Regulatory Commission, NUREG-1 560, Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance, Volume 2, December 1997.E.2-16 U.S. Nuclear Regulatory Commission, NUREG/CR-5474, Assessment of Candidate Accident Management Strategies, March 1990.E.2-17 Pilgrim Nuclear Power Station, Individual Plant Examination (IPE) Report, September 1992E.2-18 Pilgrim Nuclear Power Station, Individual Plant Examination of External Events (IPEEE)Report, July 1994.E.2-19 U.S. Nuclear Regulatory Commission, NUREG/BR-01 84, Regulatory Analysis Technical Evaluation Handbook, January 1997.QW E.2-14 t.00 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Phase II l Result of Potential SAMA ID SAMA Enhancement Improvements Related to Accident Mitigation Containment Phenomena 001 Install an SAMA would decrease 4.70% 4.60% $43,639 $261,832 $5,800,000 Not cost independent the probability of loss effective method of of containment heat suppression pool removal.cooling. I Basis for
E.24   Appendix E- Environmental Report, Appendix G, Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Peach Bottom Nuclear Power Plant Units 2 and 3, July, 2001.
E.2-5 Appendix F,Severe Accident Mitigation Alternatives Analysis Submittal Related to Licensing Renewal for the Quad Cities Nuclear Power Plant Units 1 and 2, January 2003.
E.2-6 Appendix F,Severe Accident Mitigation Alternatives Analysis Submittal Related to Licensing Renewal for the Dresden Nuclear Power Plant Units 2 and 3, January 2003.
E.2-7 Appendix E-Attachment E, Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Arkansas Nuclear One - Unit 2, October 2003.
E.2-8 Cost Estimate for Severe Accident Mitigation Design Alternatives, Limerick Generating Station for Philadelphia Electric Company, Bechtel Power Corporation, June 22, 1989.
E.2-9 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Volume 1, 5.35, Listing of SAMDAs considered for the Limerick Generating Station, May 1996.
E.2-10 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Volume 1, 5.36, Listing of SAMDAs considered for the Comanche Peak Steam Electric Station, May 1996.
E.2-11 Museler, W. J., (Tennessee Valley Authority) to NRC Document Control Desk, "Watts Bar Nuclear Plant (WBN) Units I and 2 - Severe Accident Mitigation Design Alternatives (SAMDAs)," letter dated October 7, 1994.
E.2-12 Nunn, D. E., (TVA) to NRC Document Control Desk, "Watts Bar Nuclear Plant (WBN)
Units I and 2 - Severe Accident Mitigation Design Alternatives (SAMDA) - Response to Request for Additional Information (RAI) - (TAC Nos. M77222 and M77223)," letter dated October 7, 1994.
E.2-13
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report          Pilgrim LR Proceeding SAMA Analysis                                  50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.2-13 Liparulo, N. J., (Westinghouse Electric Corporation) to NRC Document Control Desk, "Submittal of Material Pertinent to the AP600 Design Certification Review," letter dated December 15,1992.
E.2-14 U.S. Nuclear Regulatory Commission, NUREG-0498, Final Environmental Statement related to the operation of Watts Bar Nuclear Plant, Units 1 and 2, Supplement No. 1, April 1995.
E.2-15 U.S. Nuclear Regulatory Commission, NUREG-1 560, Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance, Volume 2, December 1997.
E.2-16 U.S. Nuclear Regulatory Commission, NUREG/CR-5474, Assessment of Candidate Accident Management Strategies, March 1990.
E.2-17 Pilgrim Nuclear Power Station, Individual Plant Examination (IPE) Report, September 1992 E.2-18 Pilgrim Nuclear Power Station, Individual Plant Examination of External Events (IPEEE)
Report, July 1994.
E.2-19 U.S. Nuclear Regulatory Commission, NUREG/BR-01 84, Regulatory Analysis Technical Evaluation Handbook, January 1997.
QW E.2-14
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                            Pilgrim LR Proceeding t                                                         SAMA Analysis  .00                                               50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Phase II l                         Result of Potential SAMA ID           SAMA                 Enhancement Improvements Related to Accident Mitigation Containment Phenomena 001         Install an             SAMA would decrease       4.70%           4.60%           $43,639 $261,832   $5,800,000         Not cost independent           the probability of loss                                                                           effective method of             of containment heat suppression pool       removal.
cooling.             I Basis for


== Conclusion:==
== Conclusion:==
 
The CDF contribution from loss of the torus cooling mode of RHR was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $5.8 million. Therefore, this SAMA is not cost effective for PNPS.
The CDF contribution from loss of the torus cooling mode of RHR was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $5.8 million. Therefore, this SAMA is not cost effective for PNPS.002 Install a filtered SAMA would provide 0.00% 0.00% $0 $0 $3,000,000 Not cost containment vent an alternate decay effective to provide fission heat removal method product for non-ATWS events, scrubbing.
002         Install a filtered     SAMA would provide         0.00%           0.00%           $0     $0         $3,000,000         Not cost containment vent       an alternate decay                                                                                 effective to provide fission     heat removal method product               for non-ATWS events, scrubbing.             with fission product Option 1: Gravel       scrubbing.
with fission product Option 1: Gravel scrubbing.
Bed Filter Option 2: Multiple Venturi Scrubber Basis for
Bed Filter Option 2: Multiple Venturi Scrubber Basis for


== Conclusion:==
== Conclusion:==
Successful torus venting accident progression source terms are reduced by a factor of 2 to reflect the additional filtered capability. The cost of implementing this SAMA at Peach Bottom was estimated to be $3 million. Therefore, this SAMA is not cost effective for PNPS.
E.2-15


Successful torus venting accident progression source terms are reduced by a factor of 2 to reflect the additional filtered capability.
9.~                                                                    3 NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
The cost of implementing this SAMA at Peach Bottom was estimated to be $3 million. Therefore, this SAMA is not cost effective for PNPS.E.2-15 9.~3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase II                         Result of Potential       CDF           off-Site       Estimated   Upper       Estimated BoundEsti           ated     Conclusion SAMA ID                             Enhancement         Reduction           dos           Benefit   Estimated       Cost ReductionBnet 003     Install a             Assuming that injection   0.50%           1.19%         $10,283   $61,701     >$2,000,000       Not cost containment vent     isavailable, this SAMA                                                                               effective large enough to       would provide alternate remove ATWS           decay heat removal in decay heat.           an ATWS event.
Phase II Result of Potential CDF off-Site Estimated Upper Estimated BoundEsti ated Conclusion SAMA ID Enhancement Reduction dos Benefit Estimated Cost ReductionBnet 003 Install a Assuming that injection 0.50% 1.19% $10,283 $61,701 >$2,000,000 Not cost containment vent is available, this SAMA effective large enough to would provide alternate remove ATWS decay heat removal in decay heat. an ATWS event.Basis for
Basis for


== Conclusion:==
== Conclusion:==
 
The CDF contribution from ATWS sequences associated with containment bypass were eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million.
The CDF contribution from ATWS sequences associated with containment bypass were eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million.Therefore, this SAMA is not cost effective for PNPS.004 Create a large SAMA would ensure 0.00% 48.62% $436,759 $2,620,551  
Therefore, this SAMA is not cost effective for PNPS.
>$100 million Not cost concrete crucible that molten core debris effective with heat removal escaping from the potential under vessel would be the base mat to contained within the contain molten crucible.
004     Create a large       SAMA would ensure         0.00%           48.62%         $436,759   $2,620,551   >$100 million     Not cost concrete crucible     that molten core debris                                                                             effective with heat removal     escaping from the potential under       vessel would be the base mat to       contained within the contain molten       crucible. The water core debris.         cooling mechanism would cool the molten core, preventing a melt-through of the base mat.
The water core debris. cooling mechanism would cool the molten core, preventing a melt-through of the base mat.Basis for
Basis for


== Conclusion:==
== Conclusion:==
Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at ANO-2 was estimated to be $100 million.
Therefore, this SAMA is not cost effective for PNPS.
E.2-16


Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at ANO-2 was estimated to be $100 million.Therefore, this SAMA is not cost effective for PNPS.E.2-16 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                              Pilgrim LR Proceeding SAMA Analysis                                                        50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase II SAMA Result of Potential SAMA ID Enhancement.i 005 Create a water-cooled rubble bed on the pedestal.SAMA would contain molten core debris dropping on to the pedestal and would allow the debris to be cooled.Basis for
Phase II       SAMA             Result of Potential SAMA ID                           Enhancement
                                    .i 005     Create a water-     SAMA would contain cooled rubble bed   molten core debris on the pedestal. dropping on to the pedestal and would allow the debris to be cooled.
Basis for


== Conclusion:==
== Conclusion:==
 
Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at ANO-2 was estimated to be $19 million.
Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at ANO-2 was estimated to be $19 million.Therefore, this SAMA is not cost effective for PNPS.006 Provide SAMA would provide 0.00% 0.07% $2,153 $12,915 >$1,000,000 Not cost modification for intentional flooding of effective flooding the the upper drywell head drywell head. such that if high drywell temperatures occurred, the drywell head seal would not fail.Basis for
Therefore, this SAMA is not cost effective for PNPS.
006     Provide             SAMA would provide         0.00%             0.07%         $2,153 $12,915       >$1,000,000         Not cost modification for     intentional flooding of                                                                               effective flooding the         the upper drywell head drywell head.       such that if high drywell temperatures occurred, the drywell head seal would not fail.
Basis for


== Conclusion:==
== Conclusion:==
Drywell head failures due to high temperature were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $1 million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.
E.2-17


Drywell head failures due to high temperature were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $1 million by engineering judgment.
Exhibit No. NRC000001 J                                                                  3 NRC - Applicant's Environmental Report SAMA Analysis                                                                          J Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Therefore, this SAMA is not cost effective for PNPS.E.2-17 J 3 J Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase II                         Result of Potential       CDF           Off-Site       Estimated   Upper       Estimated SAMA ID                             Enhancement         Reduction       Reduction       Benefit Estimated         Cost RedutionBenefit 007       Enhance fire         SAMA would improve       0.00%           1.16%         $10,763   $64,577       >$2,500,000         Not cost protection system   fission product                                                                                     effective and SGTS             scrubbing in severe hardware and         accidents.
Phase II Result of Potential CDF Off-Site Estimated Upper Estimated SAMA ID Enhancement Reduction Reduction Benefit Estimated Cost RedutionBenefit 007 Enhance fire SAMA would improve 0.00% 1.16%  
$10,763 $64,577 >$2,500,000 Not costprotection system fission product effective and SGTS scrubbing in severehardware and accidents.
procedures.
procedures.
Basis for
Basis for


== Conclusion:==
== Conclusion:==
Failure of the reactor building to contain releases was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2.5 million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.
E.2-18


Failure of the reactor building to contain releases was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2.5 million by engineering judgment.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                                Pilgrim LR Proceeding SAMA Analysis                                                        50-293-LR, 06-848-02-LR t_'                                                                 V                                                                     ir_'
Therefore, this SAMA is not cost effective for PNPS.E.2-18 t_' V ir_'Pilgrim Nuclear Power StationApplicant's Environmental ReportOperating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued) off-site Upper Phase II Result of Potential CDF Estimated Bound Estimated SAMA ID Enhancement Reduction Benefit Estimated Cost Conclusion ReductionBefi
Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued) off-site                 Upper Phase II                       Result of Potential         CDF                         Estimated   Bound       Estimated SAMA ID                             Enhancement         Reduction                           Benefit Estimated       Cost           Conclusion
~ : Benefit 008 Create a core melt SAMA would provide 0.00% 48.62% $436,759 $2,620,551  
                                                                        ~: ReductionBefi              Benefit 008     Create a core melt SAMA would provide         0.00%             48.62%       $436,759   $2,620,551   >$5,000,000       Not cost source reduction   cooling and                                                                                           effective system.             containment of molten core debris. Refractory material would be placed underneath the reactor vessel such that a molten core falling on the material would melt and combine with the material. Subsequent spreading and heat removal from the vitrified compound would be facilitated, and concrete attack would not occur.
>$5,000,000 Not cost source reduction cooling and effective system. containment of molten core debris. Refractory material would be placed underneath the reactor vessel such that a molten core falling on the material would melt and combine with thematerial. Subsequent spreading and heat removal from the vitrified compound would be facilitated, and concrete attack would not occur.Basis for
Basis for


== Conclusion:==
== Conclusion:==
Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $5 million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.
E.2-19


Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $5 million byengineering judgment.
3 I3__
Therefore, this SAMA is not cost effective for PNPS.E.2-19 3 I3__J, Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 J,
Phase II Result of Potential CDF OffEstimate d Upper Estimated Estmaed Bondostmaed Conclusion SAMA ID SAMA Enhancement Reduction Reduce on Benefit Estimated Cost RedutionBenefit 009 Install a passive SAMA would decrease 5.05% 4.70% $44,037 $264,219 $5,800,000 Not cost containmentspray the probability of loss effective system. of containment heat removal.Basis for
Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase II                           Result of Potential       CDF             OffEstimate   Estmaedd  Upper     Estimated Bondostmaed                 Conclusion SAMA ID         SAMA                 Enhancement         Reduction       Reduce on       Benefit Estimated       Cost RedutionBenefit 009       Install a passive       SAMA would decrease     5.05%           4.70%         $44,037   $264,219   $5,800,000         Not cost containmentspray       the probability of loss                                                                           effective system.                 of containment heat removal.
Basis for


== Conclusion:==
== Conclusion:==
 
The CDF contribution from loss of the drywell spray mode of RHR was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $5.8 million. Therefore, this SAMA is not cost effective for PNPS.
The CDF contribution from loss of the drywell spray mode of RHR was eliminated to conservatively assessthe benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $5.8 million. Therefore, this SAMAis not cost effective for PNPS.010 Strengthen SAMA would reduce 0.00% 26.10% $205,571 $1,233,428  
010       Strengthen             SAMA would reduce       0.00%           26.10%         $205,571 $1,233,428 $12,000,000         Not cost primary and             the probability of                                                                               effective secondary               containment over-containment.           pressurization failure.
$12,000,000 Not costprimary and the probability of effective secondary containment over-containment.
Basis for
pressurization failure.Basis for


== Conclusion:==
== Conclusion:==
Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR was estimated to be $12 million. Therefore, this SAMA is not cost effective for PNPS.
E.2-20


Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR was estimated to be $12 million. Therefore, this SAMA is not cost effective for PNPS.E.2-20 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                              Pilgrim LR Proceeding SAMA Analysis                                                        50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase II Result of Potential CDF lt Estimated und Estimated Estmatd BunoEsimaed Conclusion SAMA ID Enhancement Reduction Reduction Benefit Estimated Cost RedutionBenefit 011 Increase the SAMA would prevent 0.00% 0.43% $4,305 $25,831 >$5,000,000 Not cost depth of the base mat melt-through.
Phase II                         Result of Potential         CDF                 lt       Estimated Estmatd        und BunoEsimaedEstimated          Conclusion SAMA ID                           Enhancement           Reduction         Reduction         Benefit Estimated     Cost RedutionBenefit 011     Increase the         SAMA would prevent         0.00%             0.43%         $4,305     $25,831     >$5,000,000       Not cost depth of the         base mat melt-through.                                                                               effective concrete base mat or use an alternative concrete material to ensure melt-through does not occur.             l Basis for
effective concrete base mat or use an alternative concrete materialto ensure melt-through does not occur. lBasis for


== Conclusion:==
== Conclusion:==
 
Containment failure due to base mat melt-through was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $5 million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.
Containment failure due to base mat melt-through was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $5 million by engineering judgment.
012       Provide a reactor   SAMA would provide         0.00%             0.22%         $3,229     $19,373     $2,500jO00         Not cost vessel exterior       the potential to cool a                                                                               effective cooling system.       molten core before it causes vessel failure, if the lower head could be submerged in water.
Therefore, this SAMA is not cost effective for PNPS.012 Provide a reactor SAMA would provide 0.00% 0.22% $3,229 $19,373 $2,500jO00 Not cost vessel exterior the potential to cool a effective cooling system. molten core before it causes vessel failure, if the lower head couldbe submerged in water.Basis for
Basis for


== Conclusion:==
== Conclusion:==
The probability of vessel failure was modified to account for potential ex-vessel cooling of the vessel bottom head region to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $2.5 million. Therefore, this SAMA is not cost effective for PNPS.
E.2-21


The probability of vessel failure was modified to account for potential ex-vessel cooling of the vessel bottom head region to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $2.5 million. Therefore, this SAMA is not cost effective for PNPS.E.2-21 J 3j Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
J                                                                        3j NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
Phase 11 Result of Potential CDF Offite Estimated Bound Estimated Est m a ed Bo ndost m a ed C onclusion SAMA ID M Enhancement Reduction Reduction Benefit Estimated Cost Redu tionBenefit 013 Construct a SAMA would provide a 0.00% 1.16% $10,763 $64,577 >$2,000,000 Not cost building method to effective connected to depressurize primary containment and containment that reduce fission product is maintained at a release.vacuum.Basis for
Phase 11                         Result of Potential       CDF             Offite       Estimated Est ma ed   Bound Bo ndost Estimated ma ed       Conclusion SAMA ID           M                   Enhancement         Reduction       Reduction         Benefit   Estimated     Cost Redu tionBenefit 013       Construct a           SAMA would provide a     0.00%           1.16%           $10,763   $64,577   >$2,000,000         Not cost building               method to                                                                                           effective connected to           depressurize primary               containment and containment that       reduce fission product is maintained at a     release.
vacuum.
Basis for


== Conclusion:==
== Conclusion:==
 
Failure of the reactor building to contain releases was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.
Failure of the reactor building to contain releases was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.014 2.g. Dedicated SAMA would decrease 4.70% 4.60% $43,639 $261,832 $5,800,000 Not cost Suppression Pool the probability of loss effective Cooling of containment heat removal.Basis for
014       2.g. Dedicated         SAMA would decrease     4.70%           4.60%           $43,639   $261,832   $5,800,000         Not cost Suppression Pool       the probability of loss                                                                             effective Cooling               of containment heat removal.
Basis for


== Conclusion:==
== Conclusion:==
 
The CDF contribution from loss of the torus cooling mode of RHR was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $5.8 million. Therefore, this SAMA is not cost effective for PNPS.
The CDF contribution from loss of the torus cooling mode of RHR was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $5.8 million. Therefore, this SAMA is not cost effective for PNPS.015 3.a. Create a SAMA increases time 0.00% 26.10% $205,571 $1,233,428  
015       3.a. Create a         SAMA increases time     0.00%           26.10%         $205,571   $1,233,428 $8,000,000         Not cost larger volume in       before containment                                                                                 effective containment.           failure and increases time for recovery.
$8,000,000 Not cost larger volume in before containment effective containment.
Basis for
failure and increases time for recovery.Basis for


== Conclusion:==
== Conclusion:==
Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $8 million.
Therefore, this SAMA is not cost effective for PNPS.
E.2-22


Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $8 million.Therefore, this SAMA is not cost effective for PNPS.E.2-22 I:0 I A -: IK-: Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                              Pilgrim LR Proceeding I : -A I:0 SAMA Analysis                                                      50-293-LR, 06-848-02-LR IK-:
Phase II Result of Potential CDF Off-Site Uppe r Bound EstimaEsE SAMA Esimtdoond Esiatd Conclusion SAMA ID Enhancement Reduction Benefit Estimated Cost: ReductionBefi 016 3.b. Increase SAMA minimizes 0.00% 26.10%  
Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
$205,571 $1,233,428  
Phase II                         Result of Potential       CDF             Off-Site                   Upper Bound     EstimaEsE SAMA                                                                     Esimtdoond             Esiatd             Conclusion SAMA ID                             Enhancement         Reduction                           Benefit Estimated     Cost:
$12,000,000 Not cost containment likelihood of large effective pressure releases.capability (sufficient pressure to withstand severe accidents).
ReductionBefi 016     3.b. Increase         SAMA minimizes           0.00%             26.10%         $205,571   $1,233,428 $12,000,000       Not cost containment           likelihood of large                                                                                 effective pressure             releases.
capability (sufficient pressure to withstand severe accidents).
Basis for
Basis for


== Conclusion:==
== Conclusion:==
Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR was estimated to be $12 million. Therefore, this SAMA is not cost effective for PNPS.017 3.c. Install This SAMA addresses 0.00%
Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR was estimated to be $12 million. Therefore, this SAMA is not cost effective for PNPS.
0.00% $0 $0 >$1,000,000 Not cost improved vacuum the reliability of a effective breakers vacuum breaker to (redundant valves reseat following a in each line).
017     3.c. Install         This SAMA addresses       0.00%             0.00%         $0         $0         >$1,000,000       Not cost improved vacuum       the reliability of a                                                                               effective breakers             vacuum breaker to (redundant valves     reseat following a in each line).       successful opening.
successful opening.Basis for
Basis for


== Conclusion:==
== Conclusion:==
Vacuum breaker failures and suppression pool scrubbing failures were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $1 million.
Therefore, this SAMA is not cost effective for PNPS.
E.2-23


Vacuum breaker failures and suppression pool scrubbing failures were eliminated to conservatively assessthe benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $1 million.Therefore, this SAMA is not cost effective for PNPS.E.2-23 J)3 Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)O ff- ite sti m te d U p p e r Phase 11 Result of Potential CDF OffSite Estimated Und Estimated SAMA ID Enhancement Reduction Rducion Benefit Esim d Cost Redu tionBenefit 018 3.d. Increase the This SAMA would 0.00% 0.07% $2,153 $12,915 $12,000,000 Not costtemperature reduce the potential for effective margin for seals. containment failure under adverse conditions.
J)                                                                        3 NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Off- itesti m ted    Upper Phase 11                           Result of Potential         CDF             OffSite       Estimated     Und   Estimated SAMA ID                               Enhancement           Reduction       Rducion         Benefit   Esim d       Cost Redu tionBenefit 018     3.d. Increase the       This SAMA would           0.00%           0.07%           $2,153     $12,915 $12,000,000         Not cost temperature            reduce the potential for                                                                           effective margin for seals.       containment failure under adverse conditions.
Basis for
Basis for


== Conclusion:==
== Conclusion:==
 
Containment failure due to high temperature drywell seal failure was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR were estimated to be $12 million and was judged to exceed the attainable benefit, even without a detailed cost estimate. Therefore, this SAMA is not cost effective for PNPS.
Containment failure due to high temperature drywell seal failure was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR were estimated to be $12 million and was judged to exceed the attainable benefit, even without a detailed cost estimate.
019     5.b/c. Install a       SAMA would provide         0.00%           0.00%           $0         $0       $3,000,000         Not cost filtered vent           an alternate decay                                                                                 effective heat removal method for non-ATWS events, with fission product scrubbing.
Therefore, this SAMA is not cost effective for PNPS.019 5.b/c. Install a SAMA would provide 0.00% 0.00% $0 $0 $3,000,000 Not cost filtered vent an alternate decay effective heat removal method for non-ATWS events, with fission product scrubbing.
Basis for
Basis for


== Conclusion:==
== Conclusion:==
Successful torus venting accident progressions source terms are reduced by a factor of 2 to reflect the additional filtered capability. The cost of implementing this SAMA at Peach Bottom was estimated to be $3 million. Therefore, this SAMA is not cost effective for PNPS.
E.2-24


Successful torus venting accident progressions source terms are reduced by a factor of 2 to reflect theadditional filtered capability.
Exhibit No. NRC000001 NRC - Applicant's I-' Environmental Report                                Pilgrim LR Proceeding 11 delo, n SAMA Analysis                                                        50-293-LR, 06-848-02-LR 1_,
The cost of implementing this SAMA at Peach Bottom was estimated to be $3 million. Therefore, this SAMA is not cost effective for PNPS.E.2-24 I-'11 delo, n Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase II Result of Potential CDF i Estimate d BuEstimated Estmaed Bondostmaed  
Phase II                         Result of Potential         CDF                 i       Estimate Estmaedd    BuEstimated Bondostmaed                


==
== Conclusion:==
Conclusion:==


SAMA ID Enhancement Reduction De Benefit Estimated Cost Reduction Benefit 020 7.a. Provide a SAMA would provide 0.00% 0.07% $2,153 $12,915 >$1,000,000 Not cost method of drywell intentional flooding of effective head flooding.
SAMA ID                             Enhancement           Reduction             De           Benefit Estimated     Cost Reduction                   Benefit 020       7.a. Provide a       SAMA would provide         0.00%             0.07%         $2,153     $12,915     >$1,000,000       Not cost method of drywell   intentional flooding of                                                                               effective head flooding.       the upper drywell head such that if high drywell temperatures occurred, the drywell head seal would not fail.
the upper drywell head such that if high drywell temperatures occurred, the drywell head seal would not fail.Basis for
Basis for


== Conclusion:==
== Conclusion:==
 
Drywell head failures due to high temperature were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $1 million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.
Drywell head failures due to high temperature were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $1 million by engineering judgment.
021       13.a. Use           This SAMA provides         0.00%             1.16%         $10,763   $64,577     >$2,500,000       Not cost alternate method     the capability to use                                                                                 effective of reactor building firewater sprays in the spray.               reactor building to mitigate release of fission products into the reactor building following an accident.
Therefore, this SAMA is not cost effective for PNPS.021 13.a. Use This SAMA provides 0.00% 1.16% $10,763 $64,577 >$2,500,000 Not cost alternate method the capability to use effective of reactor building firewater sprays in the spray. reactor building to mitigate release of fission products into the reactor building following an accident.
Basis for
Basis for


== Conclusion:==
== Conclusion:==
Failure of the reactor building to contain releases was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2.5 million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.
E.2-25


Failure of the reactor building to contain releases was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2.5 million by engineering judgment.
3                                                                    J NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR I
Therefore, this SAMA is not cost effective for PNPS.E.2-25 3 J I Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase II Result of Potential CDF OffSte Estimated Bound Estimated Est m a ed Bo ndost m a ed C onclusion SAMA ID SAMA Enhancement Reduction Rductios Benefit Estimated Cost Redu tionBenefit 022 14.a. Provide a SAMA would allow the 0.00% 22.48% $204,495 $1,226,971  
Phase II                         Result of Potential       CDF             OffSte       Estimated Est ma ed   Bound Bo ndost Estimated ma ed       Conclusion SAMA ID         SAMA               Enhancement         Reduction       Rductios         Benefit   Estimated       Cost Redu tionBenefit 022     14.a. Provide a       SAMA would allow the     0.00%           22.48%         $204,495   $1,226,971 $2,500,000 - Not cost means of flooding     debris to be cooled.                                                                                 effective the rubble bed.
$2,500,000  
Basis for
-Not cost means of flooding debris to be cooled. effectivethe rubble bed.Basis for


== Conclusion:==
== Conclusion:==
 
The probabilities of wet core concrete interactions were substituted for dry core concrete interactions to assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $2.5 million. Therefore, this SAMA is not cost effective for PNPS.
The probabilities of wet core concrete interactions were substituted for dry core concrete interactions to assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $2.5 million. Therefore, this SAMA is not cost effective for PNPS.023 14.b. Install a SAMA would enhance 0.00% 48.62% $436,759 $2,620,551  
023     14.b. Install a       SAMA would enhance       0.00%           48.62%         $436,759   $2,620,551 $8,750,000         Not cost reactor cavity       debris coolability,                                                                                   effective flooding system.     reduce core concrete interaction, and provide fission product scrubbing.
$8,750,000 Not cost reactor cavity debris coolability, effective flooding system. reduce core concrete interaction, andprovide fission product scrubbing.
Basis for
Basis for


== Conclusion:==
== Conclusion:==
 
Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at ANO-2 was estimated to be $8.75 million.
Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at ANO-2 was estimated to be $8.75 million.Therefore, this SAMA is not cost effective for PNPS.024 Add ribbing to the This SAMA would 0.00% 26.10% $205,571 $1,233,428 $12,000,000 Not cost containment shell. reduce the chance of effective containment buckling under reverse pressure loading.Basis for
Therefore, this SAMA is not cost effective for PNPS.
024     Add ribbing to the   This SAMA would           0.00%           26.10%         $205,571   $1,233,428 $12,000,000         Not cost containment shell. reduce the chance of                                                                                 effective containment buckling under reverse pressure loading.
Basis for


== Conclusion:==
== Conclusion:==
Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR was estimated to be $12 million. Therefore, this SAMA is not cost effective for PNPS.
E.2-26


Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR was estimated to be $12 million. Therefore, this SAMA is not cost effective for PNPS.E.2-26 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                              Pilgrim LR Proceeding SAMA Analysis                                                        50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase II Result of Potential CDF Off-Site Estimated Upper Estimated Estmaed Bondostmaed Conclusion SAMA ID SAMA Enhancement Reduction Benefit Estimated Cost Improvements Related to Enhanced AC/DC Reliability/Availability R 025 Provide additional SAMA would ensure 1.39% 2.79% $24,393 $146,356 $500,000 Not cost DC battery longer battery effective capacity.
Phase II                       Result of Potential         CDF             Off-Site     Estimated Estmaed      Upper BondostmaedEstimated          Conclusion SAMA ID         SAMA             Enhancement           Reduction                           Benefit Estimated       Cost Improvements Related to Enhanced AC/DC Reliability/Availability           R 025       Provide additional SAMA would ensure         1.39%             2.79%         $24,393   $146,356   $500,000           Not cost DC battery         longer battery                                                                                       effective capacity.           capability during an SBO, which would extend HPCW/RCIC operability and allow more time for AC power recovery.
capability during anSBO, which would extend HPCW/RCIC operability and allow more time for AC power recovery.Basis for
Basis for


== Conclusion:==
== Conclusion:==
 
The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours to 24 hours during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.
The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours to 24hours during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment.
026       Use fuel cells     SAMA would extend         1.39%             2.79%         $24,393   $146,356   >$2,000,000       Not cost instead of lead-   DC power availability in                                                                             effective acid batteries. an SBO, which would extend HPCI/RCIC operability and allow more time for AC power recovery.
Therefore, this SAMA is not cost effective for PNPS.026 Use fuel cells SAMA would extend 1.39% 2.79% $24,393 $146,356 >$2,000,000 Not cost instead of lead- DC power availability in effectiveacid batteries.
Basis for
an SBO, which would extend HPCI/RCIC operability and allow more time for AC power recovery.Basis for


== Conclusion:==
== Conclusion:==
The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours to 24 hours during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2million. Therefore, this SAMA is not cost effective for PNPS.
E.2-27


The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours to 24hours during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottomwas estimated to be greater than
Exhibit No. NRC000001 J                                                      SAMA Analysis 31 NRC - Applicant's Environmental Report                                Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary     of Phase     II SAMA Candidates   Considered in Cost-Benefit Evaluation (Continued)
$2 million. Therefore, this SAMA is not cost effective for PNPS.E.2-27 J 31 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase 11                         Result of Potential         CDF           OffSite       Estimated   Bound       Estimated SAMA                                                                       Esim tdoo       ndEsi     Costatd        Conclusion SAMA ID                               Enhancement           Reduction       Rduction         Benefit   Estimated Redu tionBenefit 027       Modification for       SAMA would increase       4.65%           1.91%           $19,761   $118,568     $500,000           Not cost Improving DC Bus       reliability of AC power                                                                               effective Reliability           and injection capability.
Phase 11 Result of Potential CDF OffSite Estimated Bound Estimated SAMA Esim tdoo ndEsi atd Conclusion SAMA ID Enhancement Reduction Rduction Benefit Estimated Cost Redu tionBenefit 027 Modification for SAMA would increase 4.65% 1.91% $19,761 $118,568 $500,000 Not cost Improving DC Bus reliability of AC power effective Reliability and injection capability.
Basis for
Basis for


== Conclusion:==
== Conclusion:==
 
The CDF contribution due to loss of DC buses D16 and D07 was eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment. Therefore, this SAMAis not cost effective for PNPS.
The CDF contribution due to loss of DC buses D16 and D07 was eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment.
028       2.i. Provide 16-       SAMA includes             1.39%           2.79%           $24,393   $146,356     $500,000           Not cost hour SBO               improved capability to                                                                               effective injection.             cope with longer SBO                     l scenarios.                               l Basis for
Therefore, this SAMAis not cost effective for PNPS.028 2.i. Provide 16- SAMA includes 1.39% 2.79% $24,393 $146,356 $500,000 Not cost hour SBO improved capability to effective injection.
cope with longer SBO l scenarios.
lBasis for


== Conclusion:==
== Conclusion:==
The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours to 24 hours during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.
E.2-28


The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours to 24 hours during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                              Pilgrim LR Proceeding SAMA Analysis                                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
Therefore, this SAMA is not cost effective for PNPS.E.2-28 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
Phase II                           Result of Potential           CDF             OffEstimate   Estmaedd    Bound     Estimated Bondostmaed                 Conclusion SAMA ID       SAM                   Enhancement           Reduction                           Benefit Estimated     Cost ReductionBefi 029     9.b. Provide an         This SAMA would             2.22%             5.06%         $44,281   $265,687   >$2,000,000       Not cost alternate pump         provide a small,                                                                                     effective power source.           dedicated power source such as a dedicated diesel or gas turbine for the feedwater or condensate pumps so that they do not rely on offsite power.
Phase II Result of Potential CDF OffEstimate d Bound Estimated Estmaed Bondostmaed Conclusion SAMA ID SAM Enhancement Reduction Benefit Estimated Cost ReductionBefi 029 9.b. Provide an This SAMA would 2.22% 5.06% $44,281 $265,687 >$2,000,000 Not cost alternate pump provide a small, effective power source. dedicated power source such as a dedicated diesel or gas turbine for the feedwater or condensate pumps so that they do not rely on offsite power.Basis for
Basis for


== Conclusion:==
== Conclusion:==
 
The CDF contribution due to failure of the SBO diesel was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.
The CDF contribution due to failure of the SBO diesel was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.030 9.g. Enhance SAMA would provide 11.10% 8.47% $78,902 $473,410 $146,120 Retain procedures to increased reliability of make use of AC AC power system and bus cross-ties.
030     9.g. Enhance           SAMA would provide           11.10%           8.47%           $78,902   $473,410   $146,120           Retain procedures to           increased reliability of make use of AC         AC power system and bus cross-ties.         reduce core damage and release frequencies.
reduce core damage and release frequencies.
Basis for
Basis for


== Conclusion:==
== Conclusion:==
The CDF contribution due to loss of MCCs B17, B18, and B15 was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $146,120 by engineering judgment.
E.2-29


The CDF contribution due to loss of MCCs B17, B18, and B15 was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $146,120 by engineering judgment.E.2-29 I-)3.Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
I-)                                                                3 NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001
Off-SiteUpe Phase 11 SAMA Result of Potential CDF ose Estimated Bound Estimated SAMA ID Enhancement Reduction Benefit Estimated Cost ReductionBeet 031 10.a. Add a This SAMA addresses 24.3% 16.16% $150,504 $903,025  
                                                                                                                                              .
$3,000,000 Not cost dedicated DC the use of a diverse DC effective power supply. power system such as an additional battery orfuel cell for the purpose of providing motive power to certain components (e.g., RCIC).Basis for
Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase 11       SAMA         Result of Potential         CDF           Off-SiteUpe ose         Estimated   Bound       Estimated SAMA ID                           Enhancement           Reduction       ReductionBeet    Benefit Estimated       Cost 031       10.a. Add a       This SAMA addresses       24.3%           16.16%         $150,504 $903,025   $3,000,000         Not cost dedicated DC       the use of a diverse DC                                                                             effective power supply.     power system such as an additional battery or fuel cell for the purpose of providing motive power to certain components (e.g.,
RCIC).
Basis for


== Conclusion:==
== Conclusion:==
 
The CDF contribution due to loss of DC Bus 'B' was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $3million. Therefore, this SAMA is not cost effective for PNPS.
The CDF contribution due to loss of DC Bus 'B' was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $3 million. Therefore, this SAMA is not cost effective for PNPS.032 10.b. Install This SAMA addresses 24.3% 16.16% $150,504 $903,025 $3,000,000 Not cost additional the use of a diverse DC effective batteries or power system such as divisions.
032       10.b. Install     This SAMA addresses       24.3%           16.16%         $150,504 $903,025   $3,000,000         Not cost additional         the use of a diverse DC                                                                             effective batteries or       power system such as divisions.         an additional battery or fuel cell for the purpose of providing motive power to certain components (e.g.,
an additional battery or fuel cell for the purpose of providing motive power to certain components (e.g., RCIC).Basis for
RCIC).
Basis for


== Conclusion:==
== Conclusion:==
The CDF contribution due to loss of DC Bus 'B' was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $3 million. Therefore, this SAMA is not cost effective for PNPS.
E.2-30


The CDF contribution due to loss of DC Bus 'B' was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $3 million. Therefore, this SAMA is not cost effective for PNPS.E.2-30 VI Ar-I-Pilgrim Nuclear Power StationApplicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase I SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                              Pilgrim LR Proceeding Ar-SAMA Analysis                                                      50-293-LR, 06-848-02-LR VI                                                                                                                                       I-Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase I SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
Phase II Result of Potential CDF Off-Site Estimated Bound Estimated SAMA ID Enhancement Reduction Dose Benefit Estimated Cost Conclusion ReductionBeet Benefit 033 10.c. Install fuel SAMA would extend 1.39% 2.79% $24,393 $146,356 >$2,000,000 Not cost cells. DC power availability in effective an SBO, which would extend HPCI/RCIC operability and allow more time for AC power recovery.Basis for
Phase II                       Result of Potential         CDF             Off-Site     Estimated   Bound     Estimated SAMA ID                         Enhancement           Reduction           Dose           Benefit Estimated     Cost           Conclusion ReductionBeet Benefit 033     10.c. Install fuel SAMA would extend         1.39%             2.79%         $24,393   $146,356   >$2,000,000       Not cost cells.             DC power availability in                                                                             effective an SBO, which would extend HPCI/RCIC operability and allow more time for AC power recovery.
Basis for


== Conclusion:==
== Conclusion:==
 
The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours to 24 hours during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.
The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours to 24 hours during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.034 10.d. Enhance This SAMA would 4.65% 1.91% $19,761 $118,568 $13,000 Retain procedures to improve DC power make use of DC availability.
034     10.d. Enhance       This SAMA would           4.65%             1.91%         $19,761   $118,568   $13,000           Retain procedures to       improve DC power make use of DC     availability.
bus cross-ties.
bus cross-ties.
Basis for
Basis for


== Conclusion:==
== Conclusion:==
The CDF contribution due to loss of DC buses D16 and D17 was eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $13,000 by engineering judgment.
E.2-31


The CDF contribution due to loss of DC buses D16 and D17 was eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $13,000 by engineering judgment.
3                                                                    3 NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 3
E.2-31 3 3 3Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phas IIResut o Potntil CD Of~itUpper Phase II SAMA Result of Potential CDF Dose Estimated Bound Estimated Conclusion SAMA ID Enhancement Reduction Reduction Benefit Estimated Cost RedutionBenefit 035 10.e. Extended SAMA would extend 1.39% 2.79% $24,393 $146,356  
Phas IIResut o Potntil             CD             Of~itUpper Phase II       SAMA             Result of Potential       CDF             Dose         Estimated   Bound     Estimated           Conclusion SAMA ID                             Enhancement         Reduction       Reduction       Benefit Estimated     Cost RedutionBenefit 035       10.e. Extended       SAMA would extend         1.39%           2.79%         $24,393   $146,356   $500,000           Not cost SBO provisions.       DC power availability in                                                                           effective an SBO, which would extend HPCI/RCIC operability and allow more time for AC power recovery.
$500,000 Not cost SBO provisions.
Basis for
DC power availability in effective an SBO, which would extend HPCI/RCIC operability and allow more time for AC power recovery.Basis for


== Conclusion:==
== Conclusion:==
The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours to 24 hours during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.
Improvements in Identifying and Mitigating Containment Bypass 036      Locate RHR            SAMA would prevent        0.33%            0.21%          $2,749    $16,497    >$500,000          Not cost inside                ISLOCA outside                                                                                      effective containment.          containment.                    .
Basis for


The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours to 24 hours during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.Improvements in Identifying and Mitigating Containment Bypass 036 Locate RHR SAMA would prevent 0.33% 0.21% $2,749 $16,497
== Conclusion:==
>$500,000 Not cost inside ISLOCA outside effective containment.
RHR ISLOCA accident sequences were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be greater than $500.000. Therefore, this SAMA is not cost effective for PNPS.
containment.
037      Increase              SAMA could reduce        0.54%           0.38%           $4,025    $24,148    $100,000           Not cost frequency of valve    ISLOCA frequency.                                                                                  effective leak testing.         _                        _
.Basis for
Basis for


== Conclusion:==
== Conclusion:==
The CDF contribution due to ISLOCA was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $100,000 by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.
E.2-32


RHR ISLOCA accident sequences were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be greater than
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                              Pilgrim LR Proceeding AP"-
$500.000.
SAMA Analysis                                                        50-293-LR, 06-848-02-LR
Therefore, this SAMA is not cost effective for PNPS.037 Increase SAMA could reduce 0.54% 0.38% $4,025 $24,148 $100,000 Not cost frequency of valve ISLOCA frequency.
                                                                                                                                          , It_
effectiveleak testing.
T_
_ _Basis for
Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
Phase II                        Result of Potential        CDF            OffSit        Estimated Estmaed    Upper Bondostmaed Estimated          Conclusion SAMA ID                              Enhancement        Reduction            Dose ReductionBeet    Benefit Estimated      Cost 038      8.e. Improve          This SAMA would          0.00%             0.00%         $0        $0        >$2,000,000         Not cost MSIV design.          decrease the likelihood                                                                            effective of containment bypass scenarios.
Basis for


== Conclusion:==
== Conclusion:==
Containment bypass failure due to MSIV leakage was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.
Improvements Related to Core Cooling System 039      Install an            SAMA would allow        0.00%            0.00%          $0        $0        $135,000            Not cost independent            continued inventory in                                                                              effective diesel for the CST    CST during an SBO.
makeup pumps.
Basis for


The CDF contribution due to ISLOCA was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $100,000 by engineering judgment.
== Conclusion:==
Therefore, this SAMA is not cost effective for PNPS.E.2-32 T_AP"-, It_Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
As currently modeled, if CST water level is low, swapping HPCI/RCIC suction from the CST to the torus allows continued HPCI/RCIC injection. Therefore, the failure to switchover from CST to torus was eliminated to conservatively assess the benefit of this SAMA on CDF. The cost of implementing this SAMA was estimated to be $135,000 by engineering judgment.
Phase II Result of Potential CDF OffSit Estimated Upper Estimated Estmaed Bondostmaed Conclusion SAMA ID Enhancement Reduction Dose Benefit Estimated Cost ReductionBeet 038 8.e. Improve This SAMA would 0.00% 0.00% $0 $0 >$2,000,000 Not cost MSIV design. decrease the likelihood effective of containment bypass scenarios.
Therefore, this SAMA is not cost effective for PNPS.
E.2-33
 
9                                                                      3-NRC - Applicant's Environmental Report SAMA Analysis 3
Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase 11SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase II                           Result of Potential       CDF           Offt          Estimated Estmaed      Upped    Estimated Bondostmaed                 Conclusion SAMA ID         S                  Enhancement           Reduction           De            Benefit   Estimated     Cost ReductionBefi 040      Provide an              SAMA would reduce        3.15%           1.97%         $18,369    $110,212  >$2,000,000         Not cost additional high        frequency of core melt                                                                            effective pressure injection      from small LOCA and pump with              SBO sequences.
independent diesel.
Basis for
Basis for


== Conclusion:==
== Conclusion:==
 
The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.
Containment bypass failure due to MSIV leakage was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million.
041      Install                 SAMA would allow         3.15%           1.97%           $18,369  $110,212  >$2,000,000        Not cost independent AC          makeup capabilities                                                                                effective high pressure          during transients, small injection system.       LOCAs, and SBOs.
Therefore, this SAMA is not cost effective for PNPS.Improvements Related to Core Cooling System 039 Install an SAMA would allow 0.00% 0.00% $0 $0 $135,000 Not cost independent continued inventory in effective diesel for the CST CST during an SBO.makeup pumps.Basis for
Basis for


== Conclusion:==
== Conclusion:==
The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.
E.2-34


As currently modeled, if CST water level is low, swapping HPCI/RCIC suction from the CST to the torusallows continued HPCI/RCIC injection.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                              Pilgrim LR Proceeding SAMA Analysis                                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
Therefore, the failure to switchover from CST to torus was eliminated to conservatively assess the benefit of this SAMA on CDF. The cost of implementing this SAMA was estimated to be $135,000 by engineering judgment.Therefore, this SAMA is not cost effective for PNPS.E.2-33 9 3-3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase 11 SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Off-SiteUpper Phase 11        SAMA              Result of Potential           CDF             off-Site    Estimated     Bound    Estimated        
Phase II Result of Potential CDF Offt Estimated Upped Estimated Estmaed Bondostmaed Conclusion SAMA ID S Enhancement Reduction De Benefit Estimated Cost ReductionBefi 040 Provide an SAMA would reduce 3.15% 1.97% $18,369 $110,212 >$2,000,000 Not cost additional high frequency of core melt effective pressure injection from small LOCA and pump with SBO sequences.
independent diesel.Basis for


== Conclusion:==
== Conclusion:==


The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.041 Install SAMA would allow 3.15% 1.97% $18,369 $110,212 >$2,000,000 Not cost independent AC makeup capabilities effective high pressure during transients, small injection system. LOCAs, and SBOs.Basis for
SAMA ID                              Enhancement            Reduction            De            Benefit' Estimated    Cost ReductionBeet 042      2.a. Install a          SAMA would improve          3.15%             1.97%         $18,369     $110,212 >$2,000,000       Not cost passive high            prevention of core melt                                                                              effective pressure system.        sequences by providing additional high pressure capability to remove decay heat through an isolation condenser type system.
Basis for


== Conclusion:==
== Conclusion:==
The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.
043      2.d. Improved          SAMA will improve          2.11%            1.43%          $12,671    $76,025  >$2,000,000        Not cost high pressure          prevention of core melt                                                                              effective systems                sequences by                                                                            K improving reliability of high pressure capability to remove decay heat.
Basis for


The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.E.2-34 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
== Conclusion:==
Off-SiteUpper Phase 11 SAMA Result of Potential CDF off-Site Estimated Bound Estimated
The CDF contribution from reducing the HPCI system failure probability by a factor of 3 was estimated to bound the potential impact of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.
E.2-35


==
Exhibit No. NRC000001 a                                                                    3 NRC - Applicant's Environmental Report SAMA Analysis J
Conclusion:==
Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
Phase 11                          Result of Potential      CDF            OffSt          Estimated  Bound    Estimated SAMA                                                                      Esimtdoond          Esiatd              Conclusion SAMA ID                              Enhancement        Reduction        ReductionBeet    Benefit  Estimated    Cost 044      2.e. Install an        SAMA will improve        3.15%            1.97%          $18,369  $110,212  >$2,000,000        Not cost additional active      reliability of high-                                                                              effective high pressure          pressure decay heat system.                removal by adding an additional system.
Basis for


SAMA ID Enhancement Reduction De Benefit' Estimated Cost ReductionBeet 042 2.a. Install a SAMA would improve 3.15% 1.97% $18,369 $110,212 >$2,000,000 Not cost passive high prevention of core melt effective pressure system. sequences by providing additional high pressure capability to remove decay heat through an isolation condenser type system.Basis for
== Conclusion:==
The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.
045      8.c. Add a diverse      SAMA will improve       3.15%           1.97%           $18,369   $110,212 >$2,000,000         Not cost injection system.      prevention of core melt                                                                           effective sequences by providing additional injection capabilities.
Basis for


== Conclusion:==
== Conclusion:==
The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.
E.2-36


The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.043 2.d. Improved SAMA will improve 2.11% 1.43% $12,671 $76,025 >$2,000,000 Not cost high pressure prevention of core melt effective systems sequences by K improving reliability of high pressure capability to remove decay heat.Basis for
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                                Pilgrim LR Proceeding SAMA Analysis                                                          50-293-LR, 06-848-02-LR It---                                                                  1                                                                      It-,
Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase II 1
I Result of Potential          CDF OffSiteUpper OffSite      Estimated    Bound      Estimated            C SAMA ID          SAMA               Enhancement          Reduction            Dose          Benefit Estimated        Cost I.                               ~~~~~Reduction              Bnft    ______
Improvements Related to ATWS Mitigation 046      Increase SRV          SAMA addresses the        1.51%             0.92%         $10,600    $63,599    $2,000,000           Not cost reseat reliability. risk associated with                                                                                  effective dilution of boron caused by the failure of the SRVs to reseat after SLC injection.
Basis for


== Conclusion:==
== Conclusion:==
 
The CDF contribution due to stuck open relief valves was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.
The CDF contribution from reducing the HPCI system failure probability by a factor of 3 was estimated to bound the potential impact of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.E.2-35 a 3 J Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
047        11.a. Install an     This SAMA would          0.50%             1.19%           $10,283  $61,701    >$2,000,000           Not cost ATWS sized vent.      provide the ability to                                                                                effective remove reactor heat from ATWS events.
Phase 11 Result of Potential CDF OffSt Estimated Bound Estimated SAMA Esimtdoond Esiatd Conclusion SAMA ID Enhancement Reduction Benefit Estimated Cost ReductionBeet 044 2.e. Install an SAMA will improve 3.15% 1.97% $18,369 $110,212 >$2,000,000 Not cost additional active reliability of high- effective high pressure pressure decay heat system. removal by adding an additional system.Basis for
Basis for


== Conclusion:==
== Conclusion:==
The CDF contribution from ATWS sequences associated with containment bypass were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing of this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.
E.2-37


The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.045 8.c. Add a diverse SAMA will improve 3.15% 1.97% $18,369 $110,212 >$2,000,000 Not cost injection system. prevention of core melt effective sequences by providing additional injection capabilities.
Exhibit No. NRC000001 D
NRC - Applicant's Environmental Report SAMA Analysis  J                                                                    J Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase 11                          Result of Potential        CDF            Offit          Estimated SAMA                                                                     Esim tdoo BoundndEsi Estimated atd        Conclusion SAMA ID                            Enhancement          Reduction          Dos            Benefit Estimated      Cost ReductionBefi 048        Diversify            An alternate means of    0.00%           0.02%           $2,153    $12,915    >$200,000           Not cost explosive valve      opening a pathway to                                                                                effective operation.          the RPV for SLC system injection would improve the success probability for reactor shutdown.
Basis for
Basis for


== Conclusion:==
== Conclusion:==
 
Common cause failure of SLC explosive valves was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $200,000 by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.
The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million.
Other Improvements 049        Increase the        SAMA reduces the         0.73%           0.60%           $5,300    $31,799    >$1,500,000         Not cost reliability of SRVs  consequences of                                                                                    effective by adding signals    medium break LOCAs.
Therefore, this SAMA is not cost effective for PNPS.E.2-36 It--- 1 It-, Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued) 1 OffSiteUpper Phase II I Result of Potential CDF OffSite Estimated Bound Estimated C SAMA ID SAMA Enhancement Reduction Dose Benefit Estimated Cost I. ~~~~~Reduction Bnft ______Improvements Related to ATWS Mitigation046 Increase SRV SAMA addresses the 1.51% 0.92% $10,600 $63,599 $2,000,000 Not cost reseat reliability.
to open them automatically.
risk associated with effective dilution of boron caused by the failure of the SRVs to reseat after SLC injection.
Basis for
Basis for


== Conclusion:==
== Conclusion:==
The CDF contribution from SRVs failing to open in medium LOCA sequences was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $1.5 million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.
E.2-38


The CDF contribution due to stuck open relief valves was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $2 million at Peach Bottom. Therefore, this SAMA is not costeffective for PNPS.047 11.a. Install an This SAMA would 0.50% 1.19% $10,283 $61,701 >$2,000,000 Not cost ATWS sized vent. provide the ability to effective remove reactor heat from ATWS events.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                              Pilgrim LR Proceeding SAMA Analysis                                                        50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase II                          Result of Potential        CDF            Offsit        Estimate  d  Upper    Estimated SAMA ID        SAMA                 Enhancement          Reduction            Dose Reduction:Bnei    Benefit  Estimated    Costnclus 050      8.e. Improve SRV      This SAMA would           4.81%             3.51%           $32,396    $194,378  >$2,000,000         Not cost design.               improve SRV reliability                                                                              effective thus increasing the likelihood that sequences could be mitigated using low-pressure heat removal.
Basis for
Basis for


== Conclusion:==
== Conclusion:==
The CDF contribution from ATWS sequences associated with containment bypass were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing of this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.E.2-37 D J J Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
The probability of SRV failure to open for vessel depressurization was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.
Phase 11 Result of Potential CDF Offit Estimated Bound Estimated SAMA Esim tdoo ndEsi atd Conclusion SAMA ID Enhancement Reduction Dos Benefit Estimated Cost ReductionBefi 048 Diversify An alternate means of 0.00% 0.02% $2,153 $12,915 >$200,000 Not cost explosive valve opening a pathway to effective operation.
051      Provide self-         SAMA would eliminate      0.47%             0.55%         $4,902      $29,412  >$200,000           Not cost cooled ECCS          ECCS dependency on                                                                                    effective pump seals.           the component cooling                    ,
the RPV for SLC system injection would improve the success probability for reactor shutdown.Basis for
water system.
Basis for


== Conclusion:==
== Conclusion:==
The CDF contribution from sequences involving RHR pump failures was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $200,000 by engineering judgment.
Therefore, this SAMA is not cost effective for PNPS E.2-39


Common cause failure of SLC explosive valves was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than
Exhibit No. NRC000001 J                                                                    3 NRC - Applicant's Environmental Report SAMA Analysis Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
$200,000 by engineering judgment.
Phase II      SAMA             Result of Potential      CDF            OffSite        Estimated    Upper    Estimated Dose          Etmtd      B u dE      i aed          Conclusion SAMA ID                            Enhancement        Reduction        Reducton        Benefit  Estimated    Cost Redu tionBenefit 052      Provide digital      Upgrade plant          0.07%           0.01%         $2,352    $14,109    >$100,000           Not cost large break LOCA    instrumentation and                                                                              effective protection.         logic to improve the capability to identify symptoms/precursors of a large break LOCA (a leak before break).
Therefore, this SAMA is not cost effective for PNPS.Other Improvements 049 Increase the SAMA reduces the 0.73% 0.60% $5,300 $31,799 >$1,500,000 Not cost reliability of SRVs consequences of effectiveby adding signals medium break LOCAs.to open them automatically.
Basis for
Basis for


== Conclusion:==
== Conclusion:==
The CDF contribution due to large break LOCA was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $100,000 by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.
E.2-40


The CDF contribution from SRVs failing to open in medium LOCA sequences was eliminated toconservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                                Pilgrim LR Proceeding SAMA Analysis                                                        50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
$1.5 million by engineering judgment.
Phas IIRestto Potntil                CF          ~    iteUpper PhaseS          A            ResuMA-of Potential         CDF                         Estimated   Bound      Estimated            o SAMA ID                           Enhancement           Reduction           Dose           Benefit Estimated       Cost.           Conclusion ReductionBefi Improvements Related to IPE, IPE Update& IPEEE Insights 053      Control            This SAMA would           3.61%             2.24%         $22,873    $137,237    $300,000           Not cost containment        establish a narrow                                                                                    effective venting within a    pressure control band narrow band of      to prevent rapid pressure            containment depressurization when venting is implemented thus avoiding adverse impact on the low pressure ECCS injection systems taking suction from the torus.
Therefore, this SAMA is not cost effective for PNPS.E.2-38 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Basis for
Phase II Result of Potential CDF Offsit Estimate d Upper Estimated SAMA ID SAMA Enhancement Reduction Dose Benefit Estimated Costnclus Reduction:Bnei 050 8.e. Improve SRV This SAMA would 4.81% 3.51% $32,396 $194,378 >$2,000,000 Not cost design. improve SRV reliability effective thus increasing the likelihood that sequences could be mitigated using low-pressure heat removal.Basis for


== Conclusion:==
== Conclusion:==
The probability of SRV failure to open for vessel depressurization was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom.
The probability of the operator failing to recognize the need to vent the torus was reduced by a factor of 3 to conservatively assess the benefit of this SAMA on CDF. The cost of implementing this SAMA was estimated to be $300,000 by engineering judgment. Therefore, this SAMA Is not cost effective for PNPS.
Therefore, this SAMA is not cost effective for PNPS.051 Provide self- SAMA would eliminate 0.47% 0.55% $4,902 $29,412 >$200,000 Not cost cooled ECCS ECCS dependency on effectivepump seals.
E.2-41
the component cooling , water system.Basis for
 
J)                                                                  i NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase II                          Result of Potential      CDF                          Estimated    Upper    Estimated SAMA ID        SAMA                Enhancement        Reduction          Dose          Benefit  Estimated    Cost              onclus on ReductionBeet 054      Install a bypass      This SAMA would         0.28%           0.33%         $3,627    $21,761    $1,000,000          Not cost switch to bypass      reduce the core                                                                                    effective the low reactor        damage frequency pressure              contribution from the interlocks of LPCI    transients with stuck or core spray          open SRVs or LOCAs injection valves        cases. Core Spray and LPCI injection valves require a low permissive signal from the same two sensors to open the valves for RPV injection.
Basis for


== Conclusion:==
== Conclusion:==
 
The probability of the ECCS low-pressure permissive failing was eliminated to conservatively assess the benefit of this SAMA on CDF. The cost of implementing this SAMA at Dresden was estimated to be $1 million. Therefore, this SAMA is not cost effective for PNPS.
The CDF contribution from sequences involving RHR pump failures was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $200,000 by engineering judgment.
055      Increase the          This SAMA would        4.37%           6.63%           $59,385  $356,310  >$5 million        Not cost reliability of SSW      reduce common cause                                                                              effective and RBCCW              dependencies from pumps.                  SSW and RBCCW systems and thus reduce plant risk.
Therefore, this SAMA is not cost effective for PNPS E.2-39 J 3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
Basis for
Phase II Result of Potential CDF OffSite Estimated Upper Estimated SAMA Dose Etmtd B u dE i aed Conclusion SAMA ID Enhancement Reduction Reducton Benefit Estimated Cost Redu tionBenefit 052 Provide digital Upgrade plant 0.07% 0.01% $2,352 $14,109 >$100,000 Not costlarge break LOCA instrumentation and effective protection.
logic to improve the capability to identify symptoms/precursors of a large break LOCA (a leak before break).Basis for


== Conclusion:==
== Conclusion:==
The CDF contribution from sequences involving common cause failures of SSW and RBCCW was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $5million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.
E.2-42


The CDF contribution due to large break LOCA was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $100,000 by engineering judgment.
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                              Pilgrim LR Proceeding SAMA Analysis                                                      50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
Therefore, this SAMA is not cost effective for PNPS.E.2-40 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase II        SAMA            Result of Potential         CDF             Off-Site      Estimated   Upper    Estimated SAMA ID         S                  Enhancement           Reduction        Reduction         Benefit Estimated     Cost RedutionBenefit 056      Provide redundant    This SAMA would             8.81%             3.51%         $36,773    $220,639  $112,400          Retain DC power            improve reliability of supplies to DTV      the DTV valves and valves.              enhance containment heat removal capability.
Phas IIRestto Potntil CF ~ iteUpper PhaseS A ResuMA-of Potential CDF Estimated Bound Estimated o SAMA ID Enhancement Reduction Dose Benefit Estimated Cost. Conclusion ReductionBefi Improvements Related to IPE, IPE Update& IPEEE Insights 053 Control This SAMA would 3.61% 2.24% $22,873 $137,237 $300,000 Not costcontainment establish a narrow effective venting within a pressure control band narrow band of to prevent rapid pressure containment depressurization when venting is implemented thus avoiding adverse impact on the low pressure ECCS injection systems taking suction from the torus.Basis for
Basis for


== Conclusion:==
== Conclusion:==
 
The CDF contribution from sequences involving DC power supply failures to the DTV valves was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $112,400 by engineering judgment.
The probability of the operator failing to recognize the need to vent the torus was reduced by a factor of 3 to conservatively assess the benefit of this SAMA on CDF. The cost of implementing this SAMA was estimated to be $300,000 byengineering judgment.
057      Proceduralize use    This SAMAwould              2.25%             3.14%           $29,213  $175,279    $26,000           Retain of the diesel fire    increase capability to pump hydro            provide makeup to the turbine in the       fire pump- day tank to event of EDG A      allow continued failure or           operation of the diesel unavailability.     fire pump, without dependence on electrical power.
Therefore, this SAMA Is not cost effective for PNPS.E.2-41 J)i Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Phase II Result of Potential CDF Estimated Upper Estimated SAMA ID SAMA Enhancement Reduction Dose Benefit Estimated Cost onclus on ReductionBeet 054 Install a bypass This SAMA would 0.28% 0.33% $3,627 $21,761 $1,000,000 Not cost switch to bypass reduce the core effective the low reactor damage frequency pressure contribution from the interlocks of LPCI transients with stuck or core spray open SRVs or LOCAs injection valves cases. Core Spray and LPCI injection valves require a low permissive signal fromthe same two sensors to open the valves for RPV injection.
Basis for
Basis for


== Conclusion:==
== Conclusion:==
The CDF contribution from sequences involving a LOOP and failure of either EDG A, or the EDG A fuel oil transfer oil pump, was eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be
          $26,000 by engineering judgment.
E.2-43


The probability of the ECCS low-pressure permissive failing was eliminated to conservatively assess the benefit of this SAMA on CDF. The cost of implementing this SAMA at Dresden was estimated to be $1 million. Therefore, this SAMA is not cost effective for PNPS.055 Increase the This SAMA would 4.37% 6.63% $59,385 $356,310
3                                                                    3 NRC - Applicant's Environmental Report SAMA Analysis 3
>$5 million Not cost reliability of SSW reduce common cause effective and RBCCW dependencies from pumps. SSW and RBCCW systems and thus reduce plant risk.Basis for
Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
Upper Phase 11                          Result of Potential        CDF             f-eeneit                Buppenr      Cst SAMA ID        SAMA                 Enhancement          Reduction          Dose          Benefit  Estimated    Cost            Cnlso 058      Proceduralize the     This SAMA would           4.92%           3.14%         $31,799  $190,797  $50,000            Retain operator action to    provide the direction to feed BI loads via    restore B15 and B17 B3 When A5 is        loads upon loss of A5 unavailable post-    initiating events as long trip. Similarly,      as A3 is available.
feed B2 loads via    Additionally, it would B4 when A6 is        provide the direction to unavailable post      restore B14 and B18 trip.,                loads upon loss of A6 initiating events as long as A4 is available.
Basis for


== Conclusion:==
== Conclusion:==
 
The CDF contribution from sequences involving loss of 4160VAC safeguard bus AS was conservatively eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $50,000 by engineering judgment.
The CDF contribution from sequences involving common cause failures of SSW and RBCCW was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $5 million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.E.2-42 Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
059      Provide redundant     This SAMA would           8.77%           17.19%         $154,966  $929,797  $1,956,000          Not cost path from fire        enhance the                                                                                        effective protection pump      availability and discharge to LPCI    reliability of the loops A and B        firewater cross-tie to cross-tie.           LPCI loops A and B for reactor vessel injection and drywell spray.
Phase II SAMA Result of Potential CDF Off-Site Estimated Upper Estimated SAMA ID S Enhancement Reduction Reduction Benefit Estimated Cost RedutionBenefit 056 Provide redundant This SAMA would 8.81% 3.51%  
$36,773 $220,639 $112,400 Retain DC power improve reliability of supplies to DTV the DTV valves and valves. enhance containment heat removal capability.
Basis for
Basis for


== Conclusion:==
== Conclusion:==
The CDF contribution from sequences involving DC power supply failures to the DTV valves was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $112,400 by engineering judgment.057 Proceduralize use This SAMAwould 2.25% 3.14% $29,213 $175,279 $26,000 Retain of the diesel fire increase capability to pump hydro provide makeup to the turbine in the fire pump- day tank to event of EDG A allow continuedfailure or operation of the diesel unavailability.
The CDF contribution from sequences involving firewater injection failures was conservatively eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $1,956,000 by engineering judgment.
fire pump, without dependence on electrical power.Basis for
Therefore, this SAMA isnot cost effective for PNPS E.2-44
 
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                                    Pilgrim LR Proceeding SAMA Analysis                                                            50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results Upper                                            Upper                              Upper PhEstimated                                      Bound                              Estimaited    Bound        Estimated            Bound II                                  Benefit  Estimated                              Benefit    Estimated        Benefit          Estimated SAMASAMA                                          Benefit              Cost                        Benefit                            Benefit IDBase                                    Line Base Lne                            Sensitivity  Sensitivity    Sensitivity        Sensitivity Case I      Case I          Case 2            Case 2 I    Install an independent          $43,639    $261,832        $5,800,000         $50,320      $301,920        $59,355          $356,129 method of suppression pool cooling.
2    Install a filtered containment        $0            $0      $3,000,000                $0          $0                $0              $0 vent to provide fission product scrubbing. Option 1: Gravel Bed Filter Option 2: Multiple Venturi Scrubber 3    Install a containment vent      $10,283    $61,701        >$2,000,000          $11,702      $70,211        $14,207          $85,244 large enough to remove ATWS decay heat.
4    Create a large concrete        $436,759  $2,620,551      >$100 million        $492,136    $2,952,813        $610,307        $3,661,845 crucible with heat removal potential under the basemat to contain molten core debris.
5    Create a water-cooled rubble    $436,759  $2,620,551        $19,000,000        $498,057    $2,988,339        $610,307        $3,661,845 bed on the pedestal.
E.2-45
 
3                                                              J NRC - Applicant's Environmental Report SAMA Analysis
                                                                                                                                            .a..
Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
Upper                                            Upper                            Upper Phase                                Estimated      Bound                              Estimated    Bound        Estimated          Bound II                                  Benefit    Estimated        Estimated          Benefit    Estimated        Benefit        Estimated SAASAMA                                        Benefit            CotBenefit                                                    Benefit ID                                                                                  Sensitivity  Sensitivity    Sensitivity      Sensitivity BsLie BeLieCase                                    I    Case I          Case 2          Case 2 6    Provide modification for          $2,153      $12,915      >$1,000,000              $2,425    $14,551          $3,008          $18,048 flooding the drywell head 7    Enhance fire protection        $10,763        $64,577      >$2,500,000            $12,127      $72,764        $15,040          $90,238 system and/or SGTS hardware and procedures.
8    Create a core melt source      $436,759    $2,620,551      >$5,000,000          $498,057    $2,988,339        $610,307      $3,661,845 reduction system.
9    Install a passive containment  $44,037      $264,219        $5,800,000          $50,845    $305,069        $59,803        $358,816 spray system.,
10    Strengthen primary/            $205,571    $1,233,428      $12,000,000          $231,636    $1,389,815        $287,257      $1,723,540 secondary containment.
11    Increase the depth of the         $4,305      $25,831      >$5,000,000            $4,851    $29,105          $6,016          $36,095 concrete basemat or use an alternative concrete material to ensure melt-through does not occur 12    Provide a reactor vessel          $3,229      $19,373        $2,500,000            $3,638    $21,828          $4,512          $27,071 exterior cooling system (see
      #7)                                                                                      _
E.2-46


== Conclusion:==
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                                    Pilgrim I- LR 1 Proceeding A4*1 SAMA Analysis                                                            50-293-LR, 06-848-02-LR 1--
Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
Upper                                            Upper                              Upper Phase                                Estimated    Bound                            Estimated      Bound        Estimated            Bound II                                  Benefit    Estimated                              Benefit    Estimated        Benefit          Estimated SAMA                SAMA                            Benefit              Cost                        Benefit                            Benefit ID                                      L                                          Sensitivity  Sensitivity    Sensitivity        Sensitivity Base                                        Case I      Case I          Case 2            Case 2 13  Construct a building to be        $10,763      $64,577        >$2,000,000          $12,273      $73,640        $15,040            $90,238 connected to primary/
secondary containment that is maintained at a vacuum 14  2.g. Dedicated Suppression        $43,639      $261,832        $5,800,000          $51,067      $306,400        $59,355          $356,129 Pool Cooling 15  3.a. Create a larger volume in  $205,571    $1,233,428        $8,000,000        $234,423    $1,406,537        $287,257        $1,723,540 containment.
16  3.b. Increase containment        $205,571    $1,233,428        $12,000,000        $234,423    $1,406,537        $287,257        $1,723,540 pressure capability (sufficient pressure to withstand severe accidents).
17  3.c. Install improved vacuum            $0            $o      >$1,000,000                $0          $0                $0              $0 breakers (redundant valves in each line).
18  3.d. Increase the temperature      $2,153      $12,915        $12,000,000          $2,455      $14,728          $3,008          $18,048 margin for seals.
19  5.b/c. install a filtered vent          $0            $0        $3,000,000                $0          $0                $0              $0 E.2-47


The CDF contribution from sequences involving a LOOP and failure of either EDG A, or the EDG A fuel oil transfer oil pump, was eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be$26,000 by engineering judgment.E.2-43 3 3 3 Pilgrim Nuclear Power Station Applicant's Environmental ReportOperating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
    -3Ad                                                            U NRC - Applicant's Environmental Report SAMA Analysis 9,
Upper Phase 11 Result of Potential CDF f-eeneit Buppenr Cst SAMA ID SAMA Enhancement Reduction Dose Benefit Estimated Cost Cnlso 058 Proceduralize the This SAMA would 4.92% 3.14% $31,799 $190,797 $50,000 Retain operator action to provide the direction to feed BI loads via restore B15 and B17 B3 When A5 is loads upon loss of A5 unavailable post- initiating events as long trip. Similarly, as A3 is available.
Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
feed B2 loads via Additionally, it would B4 when A6 is provide the direction to unavailable post restore B14 and B18 trip., loads upon loss of A6initiating events as long as A4 is available.
Upper                                          Upper                              Upper Phase                                 Estimated      Bound                            Estimated      Bound        Estimated            Bound 11                                   Benefit    Estimated          Estimated          Benefit    Estimated          Benefit        Estimated SAMA                             Benefit             Cost                       Benefit                            Benefit ID                                                                                    Sensitivity  Sensitivity    Sensitivity      Sensitivity Case I      Case I          Case 2            Case 2 20    7.a. Provide a method of            $2,153      $12,915      >$1,000,000            $2,455    $14,728          $3,008          $18,048 drywell head flooding.
Basis for
21    13.a. Use alternate method of    $10,763        $64,577      >$2,500,000            $12,273    $73,640        $15,040          $90,238 reactor building spray.
22    14.a. Provide a means of        $204,495    $1,226,971        $2,500,000        $230,423  $1,382,539        $285,753        $1,714,516 flooding the rubble bed.
23    14.b. Install a reactor cavity  $436,759    $2,620,551        $8,750,000         $498,057  $2,988,339        $610,307        $3,661,845 flooding system.
24    Add ribbing to the               $205,571    $1,233,428      $12,000,000          $234,423  $1,406,537        $287,257        $1,723,540 containment shell.
25    Provide additional DC battery    $24,393      $146,356          $500,000          $27,830    $166,978        $33,598          $201,588 capacity.
26    Use fuel cells instead of lead-   $24,393      $146,356      >$2,000,000            $28,207    $169,242        $33,598          $201,588 acid batteries.
27    Modification for Improving DC    $19,761      $118,568          $500,000          $23,377    $140,262        $26,044          $156,263 Bus Reliability 28  2.i. Provide 16-hour SBO          $24,393      $146,356          $500,000          $28,207    $169,242        $33,598          $201,588 injection.
E.2-48


== Conclusion:==
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                                    Pilgrim LR Proceeding SAMA Analysis                                                              50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
Upper                                              Upper                              Upper Phase                            Estimated    Bound                              Estimated      Bound        Estimated            Bound IP                              Benefit    Estimated          Estimated          Benefit    Estimated          Benefit        Estimated SSAMA,                                          Benefit              Cost                          Benefit                            Benefit ID                                                                              Sensitivity    Sensitivity    Sensitivity      Sensitivity Base Line  Base Line                              Case I    -  Case 1          Case 2            Case 2 29  9.b. Provide an alternate    $44,281      $265,687        >$2,000,000          $50,546      $303,278        $60,956          $365,738 pump power source.
30  9.g. AC Bus Cross-Ties        $78,902    $473,410            $146,120        $91,662      $549,972        $106,357          $638,142 31  10.a. Add a dedicated DC    $150,504    $903,025          $3,000,000        $178,405    $1,070,432        $201,864        $1,211,183 power supply.
32  10.b. Install additional    $150,504    $903,025          $3,000,000        $178,405    $1,070,432        $201,864        $1,211,183 batteries or divisions.
33  10.c. Install fuel cells.    $24,393      $146,356        >$2,000,000          $28,207      $169,242          $33,598        $201,588 34  10.d. DC Cross-Ties          $19,761      $118,568            $13,000        $23,377      $140,262        $26,044          $156,263 35  10.e. Extended SBO            $24,393      $146,356            $500,000        $28,207      $169,242        $33,598          $201,588 provisions.
36  Locate RHR inside              $2,749      $16,497          >$500,000          $3,213        $19,276          $3,680          $22,077 containment.
37  Increase frequency of valve    $4,025      $24,148            $100,000          $4,688        $28,127          $5,407          $32,444 leak testing.
38  8.e. improve MSIV design.          $0            $0      >$2,000,000                $0            $0                $0              $0 E.2-49


The CDF contribution from sequences involving loss of 4160VAC safeguard bus AS was conservatively eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $50,000 by engineering judgment.059 Provide redundant This SAMA would 8.77% 17.19% $154,966 $929,797 $1,956,000 Not costpath from fire enhance the effective protection pump availability and discharge to LPCI reliability of the loops A and B firewater cross-tie to cross-tie.
3 NRC - Applicant's Environmental Report SAMA Analysis                                                                            a Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
LPCI loops A and B forreactor vessel injection and drywell spray.Basis for
Upper                                            Upper                              Upper Phase                              Estimated    Bound                            Estimated    Bound          Estimated            Bound Ii                                Benefit    Estimated                            Benefit    Estimated          Benefit        Estimated SAMA  SAMA                                       Benefit            Cost                        Benefit                            Benefit ID                                    ; L                                        Sensitivity  Sensitivity    Sensitivity      Sensitivity Base Line  Base Line                            Case I      Case I            Case 2            Case 2 39  Install an independent diesel        $0            $0          $135,000               $0          $0                $0              $0 for the CST makeup pumps.
40  Provide an additional high      $18,369      $110,212      >$2,000,000            $21,540    $129,238        $24,477          $146,860 pressure injection pump with independent diesel.
41  Install independent AC high    $18,369      $110,212      >$2,000,000            $21,902    $131,415          $24,477        $146,860 pressure injection system.
42  2.a. Install a passive high    $18,369      $110,212      >$2,000,000            $21,902    $131,415          $24,477        $146,860 pressure system.
43  2.d. Improved high pressure    $12,671      $76,025      >$2,000,000            $14,851    $89,109          $16,894        $101,363 systems 44  2.e. Install an additional      $18,369      $110,212      >$2,000,000            $21,902    $131,415          $24,477        $146,860 active high pressure system.
45  8.c. Add a diverse injection    $18,369      $110,212      >$2,000,000            $21,902    $131,415          $24,477        $146,860 system.
46  Increase SRV reseat            $10,600      $63,599        $2,000,000           $12,326    $73,958          $14,270          $85,623 reliability.
47  11.a. Install an ATWS sized    $10,283      $61,701      >$2,000,000            $11,857    $71,142          $14,207          $85,244 vent.
E.2-50


== Conclusion:==
Exhibit No. NRC000001 NRC - Applicant's Environmental Report                                  Pilgrim LR Proceeding SAMA Analysis                                                          50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
Upper                                          Upper                              Upper Phase                                Estimated    Bound                            Estimated    Bound          Estimated            Bound 11                                  Benefit    Estimated          Estimated          Benefit  Estimated        Benefit        Estimated SSAMA                                              Benefit            Cost                      Benefit                            Benefit ID                                                                                  Sensitivity Sensitivity    Sensitivity      Sensitivity Base Line  Base Line                              Case I    Case I          Case 2            Case 2 48  Diversify explosive valve          $2,153      $12,915        >$200,000            $2,425    $14,551          $3,008          $18,048 operation.
49  Increase the reliability of        $5,300      $31,799        >$1,500,000          $6,163    $36,978          $7,135          $42,811 SRVs by adding signals to open them automatically.
50  8.e. Improve SRV design.        $32,396    $194,378        >$2,000,000          $37,767    $226,602          $43,483        $260,897 51  Provide self-cooled ECCS          $4,902      $29,412        >$200,000            $5,638    $33,829          $6,687          $40,125 pump seals.
52  Provide digital large break        $2,352      $14,109          >$1 00,000          $2,688    $16,126          $3,232          $19,391 LOCA protection.
53  Control containment venting      $22,873      $137,237            $300,000        $26,653    $159,919          $30,716        $184,299 within a narrow band of pressure 54  Install a bypass switch to        $3,627      $21,761        $1,000,000          $4,163    $24,978          $4,960          $29,758 bypass the low reactor pressure interlocks of LPCI or core spray injection valves.
55  Improve SSW System and          $59,385      $356,310        >$5 million        $67,986    $407,918          $81,467        $488,799 RBCCW pump recovery.
E.2-51


The CDF contribution from sequences involving firewater injection failures was conservatively eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $1,956,000 by engineering judgment.Therefore, this SAMA is not cost effective for PNPS E.2-44 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results Upper Upper Upper PhEstimated Bound Estimaited Bound Estimated Bound II Benefit Estimated Benefit Estimated Benefit Estimated SAMASAMA Benefit Cost Benefit Benefit IDBase Line Base Lne Sensitivity Sensitivity Sensitivity Sensitivity Case I Case I Case 2 Case 2 I Install an independent
J                                                             3 NRC - Applicant's Environmental Report SAMA Analysis
$43,639 $261,832 $5,800,000
                                                                                                                                            .,)
$50,320 $301,920 $59,355 $356,129 method of suppression pool cooling.2 Install a filtered containment
Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
$0 $0 $3,000,000
Upper                                           Upper                               Upper Phase                               Estimated     Bound                             Estimated     Bound         Estimated           Bound II                               Benefit   Estimated         Estimated         Benefit   Estimated         Benefit         Estimated SAMA               SAMA                           Benefit             Cost                       Benefit                             Benefit ID                                                                               Sensitivity Sensitivity     Sensitivity       Sensitivity Base Line   Base Line                             Case I       Case I           Case 2           Case 2 56  Provide redundant DC power     $36,773     $220,639           $112,400           $43,541   $261,247         $48,408         $290,449 supplies to DTV valves.
$0 $0 $0 $0 vent to provide fission product scrubbing.
57 Proceduralize the use of       $29,213     $175,279           $26,000           $33,568   $201,406         $39,901         $239,406 diesel fire pump hydroturbine in the event of EDG A failure or unavailability.
Option 1: Gravel Bed Filter Option 2: Multiple Venturi Scrubber 3 Install a containment vent $10,283 $61,701 >$2,000,000
58 Proceduralize the operator     $31,799     $190,797             $50,000         $36,980   $221,878         $42,811         $256,868 action to feed B1loads via B3 When AS is unavailable post-trip.
$11,702 $70,211 $14,207 $85,244 large enough to remove ATWS decay heat.4 Create a large concrete $436,759 $2,620,551
59 Provide redundant path from   $154,966     $929,797         $1,956,000         $176,682   $1,060,091       $213,620       $1,281,720 fire protection pump discharge to LPCI loops A and B cross-tie.
>$100 million $492,136 $2,952,813
$610,307 $3,661,845 crucible with heat removal potential under the basemat to contain molten core debris.5 Create a water-cooled rubble $436,759 $2,620,551
$19,000,000
$498,057 $2,988,339
$610,307 $3,661,845 bed on the pedestal.E.2-45 3 J.a..Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound II Benefit Estimated Estimated Benefit Estimated Benefit Estimated SAASAMA Benefit CotBenefit Benefit ID Sensitivity Sensitivity Sensitivity Sensitivity BsLie BeLieCase I Case I Case 2 Case 2 6 Provide modification for $2,153 $12,915 >$1,000,000
$2,425 $14,551 $3,008 $18,048 flooding the drywell head 7 Enhance fire protection
$10,763 $64,577 >$2,500,000
$12,127 $72,764 $15,040 $90,238 system and/or SGTS hardware and procedures.
8 Create a core melt source $436,759 $2,620,551
>$5,000,000
$498,057 $2,988,339
$610,307 $3,661,845 reduction system.9 Install a passive containment
$44,037 $264,219 $5,800,000
$50,845 $305,069 $59,803 $358,816 spray system., 10 Strengthen primary/ $205,571 $1,233,428
$12,000,000
$231,636 $1,389,815
$287,257 $1,723,540 secondary containment.
11 Increase the depth of the $4,305 $25,831 >$5,000,000
$4,851 $29,105 $6,016 $36,095 concrete basemat or use an alternative concrete material to ensure melt-through does not occur 12 Provide a reactor vessel $3,229 $19,373 $2,500,000
$3,638 $21,828 $4,512 $27,071 exterior cooling system (see#7) _E.2-46 I- 1 A4*1Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound II Benefit Estimated Benefit Estimated Benefit Estimated SAMA SAMA Benefit Cost Benefit Benefit ID L Sensitivity Sensitivity Sensitivity Sensitivity Base Case I Case I Case 2 Case 2 13 Construct a building to be $10,763 $64,577 >$2,000,000
$12,273 $73,640 $15,040 $90,238 connected to primary/secondary containment that is maintained at a vacuum 14 2.g. Dedicated Suppression
$43,639 $261,832 $5,800,000
$51,067 $306,400 $59,355 $356,129 Pool Cooling 15 3.a. Create a larger volume in $205,571 $1,233,428
$8,000,000
$234,423 $1,406,537
$287,257 $1,723,540 containment.
16 3.b. Increase containment
$205,571 $1,233,428
$12,000,000
$234,423 $1,406,537
$287,257 $1,723,540 pressure capability (sufficient pressure to withstand severe accidents).
17 3.c. Install improved vacuum $0 $o >$1,000,000
$0 $0 $0 $0 breakers (redundant valves in each line).18 3.d. Increase the temperature
$2,153 $12,915 $12,000,000
$2,455 $14,728 $3,008 $18,048 margin for seals.19 5.b/c. install a filtered vent $0 $0 $3,000,000
$0 $0 $0 $0 E.2-47
-3Ad U 9, Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound 11 Benefit Estimated Estimated Benefit Estimated Benefit Estimated SAMA Benefit Cost Benefit Benefit ID Sensitivity Sensitivity Sensitivity Sensitivity Case I Case I Case 2 Case 2 20 7.a. Provide a method of $2,153 $12,915 >$1,000,000
$2,455 $14,728 $3,008 $18,048 drywell head flooding.21 13.a. Use alternate method of $10,763 $64,577 >$2,500,000
$12,273 $73,640 $15,040 $90,238reactor building spray.22 14.a. Provide a means of $204,495 $1,226,971 $2,500,000 $230,423
$1,382,539
$285,753 $1,714,516flooding the rubble bed.23 14.b. Install a reactor cavity $436,759 $2,620,551
$8,750,000
$498,057 $2,988,339
$610,307 $3,661,845 flooding system.24 Add ribbing to the $205,571 $1,233,428
$12,000,000
$234,423 $1,406,537
$287,257 $1,723,540 containment shell.25 Provide additional DC battery $24,393 $146,356 $500,000 $27,830 $166,978 $33,598 $201,588 capacity.26 Use fuel cells instead of lead- $24,393 $146,356 >$2,000,000 $28,207 $169,242
$33,598 $201,588 acid batteries.
27 Modification for Improving DC $19,761 $118,568 $500,000 $23,377 $140,262 $26,044 $156,263 Bus Reliability 28 2.i. Provide 16-hour SBO $24,393 $146,356 $500,000 $28,207 $169,242
$33,598 $201,588 injection.
E.2-48 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound IP Benefit Estimated Estimated Benefit Estimated Benefit Estimated SSAMA, Benefit Cost Benefit Benefit ID Sensitivity Sensitivity Sensitivity Sensitivity Base Line Base Line Case I -Case 1 Case 2 Case 2 29 9.b. Provide an alternate
$44,281 $265,687 >$2,000,000
$50,546 $303,278 $60,956 $365,738 pump power source.30 9.g. AC Bus Cross-Ties
$78,902 $473,410 $146,120 $91,662 $549,972 $106,357 $638,142 31 10.a. Add a dedicated DC $150,504 $903,025 $3,000,000
$178,405 $1,070,432
$201,864 $1,211,183 power supply.32 10.b. Install additional
$150,504 $903,025 $3,000,000
$178,405 $1,070,432
$201,864 $1,211,183 batteries or divisions.
33 10.c. Install fuel cells. $24,393 $146,356 >$2,000,000
$28,207 $169,242 $33,598 $201,588 34 10.d. DC Cross-Ties
$19,761 $118,568 $13,000 $23,377 $140,262 $26,044 $156,263 35 10.e. Extended SBO $24,393 $146,356 $500,000 $28,207 $169,242 $33,598 $201,588 provisions.
36 Locate RHR inside $2,749 $16,497 >$500,000
$3,213 $19,276 $3,680 $22,077 containment.
37 Increase frequency of valve $4,025 $24,148 $100,000 $4,688 $28,127 $5,407 $32,444 leak testing.38 8.e. improve MSIV design. $0 $0 >$2,000,000
$0 $0 $0 $0 E.2-49 3 a Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound Ii Benefit Estimated Benefit Estimated Benefit Estimated SAMA SAMA Benefit Cost Benefit Benefit ID ; L Sensitivity Sensitivity Sensitivity Sensitivity Base Line Base Line Case I Case I Case 2 Case 2 39 Install an independent diesel $0 $0 $135,000 $0 $0 $0 $0 for the CST makeup pumps.40 Provide an additional high $18,369 $110,212 >$2,000,000
$21,540 $129,238 $24,477 $146,860pressure injection pump with independent diesel.41 Install independent AC high $18,369 $110,212 >$2,000,000
$21,902 $131,415 $24,477 $146,860pressure injection system.42 2.a. Install a passive high $18,369 $110,212 >$2,000,000
$21,902 $131,415 $24,477 $146,860 pressure system.43 2.d. Improved high pressure $12,671 $76,025 >$2,000,000
$14,851 $89,109 $16,894
$101,363 systems 44 2.e. Install an additional
$18,369 $110,212 >$2,000,000
$21,902 $131,415 $24,477 $146,860 active high pressure system.45 8.c. Add a diverse injection
$18,369 $110,212 >$2,000,000
$21,902 $131,415 $24,477 $146,860 system.46 Increase SRV reseat $10,600 $63,599 $2,000,000
$12,326 $73,958 $14,270
$85,623 reliability.
47 11.a. Install an ATWS sized $10,283 $61,701 >$2,000,000
$11,857 $71,142 $14,207
$85,244 vent.E.2-50 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound 11 Benefit Estimated Estimated Benefit Estimated Benefit Estimated SSAMA Benefit Cost Benefit Benefit ID Sensitivity Sensitivity Sensitivity Sensitivity Base Line Base Line Case I Case I Case 2 Case 2 48 Diversify explosive valve $2,153 $12,915 >$200,000
$2,425 $14,551 $3,008 $18,048 operation.
49 Increase the reliability of $5,300 $31,799 >$1,500,000
$6,163 $36,978 $7,135 $42,811 SRVs by adding signals to open them automatically.
50 8.e. Improve SRV design. $32,396 $194,378 >$2,000,000
$37,767 $226,602 $43,483 $260,897 51 Provide self-cooled ECCS $4,902 $29,412 >$200,000
$5,638 $33,829 $6,687 $40,125 pump seals.52 Provide digital large break $2,352 $14,109 >$1 00,000 $2,688 $16,126 $3,232 $19,391 LOCA protection.
53 Control containment venting $22,873 $137,237 $300,000 $26,653 $159,919 $30,716 $184,299 within a narrow band of pressure 54 Install a bypass switch to $3,627 $21,761 $1,000,000
$4,163 $24,978 $4,960 $29,758 bypass the low reactor pressure interlocks of LPCI or core spray injection valves.55 Improve SSW System and $59,385 $356,310 >$5 million $67,986 $407,918 $81,467 $488,799 RBCCW pump recovery.
E.2-51 J 3.,)Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound II Benefit Estimated Estimated Benefit Estimated Benefit Estimated SAMA SAMA Benefit Cost Benefit Benefit ID Sensitivity Sensitivity Sensitivity Sensitivity Base Line Base Line Case I Case I Case 2 Case 256 Provide redundant DC power $36,773 $220,639 $112,400 $43,541 $261,247 $48,408 $290,449 supplies to DTV valves.57 Proceduralize the use of $29,213 $175,279 $26,000 $33,568 $201,406 $39,901 $239,406 diesel fire pump hydroturbine in the event of EDG A failure or unavailability.
58 Proceduralize the operator $31,799  
$190,797 $50,000 $36,980 $221,878 $42,811 $256,868action to feed B1 loads via B3 When AS is unavailable post-trip.59 Provide redundant path from $154,966 $929,797 $1,956,000  
$176,682 $1,060,091  
$213,620 $1,281,720fire protection pump discharge to LPCI loops A and B cross-tie.
E.2-52}}
E.2-52}}

Revision as of 03:41, 13 November 2019

Official Exhibit - NRC000001-00-BD01 - Applicant'S Environmental Report, Attachment E, Severe Accident Mitigation Alternatives Analysis.
ML110600910
Person / Time
Site: Pilgrim
Issue date: 01/03/2011
From:
NRC/OGC
To:
Atomic Safety and Licensing Board Panel
SECY RAS
Shared Package
ML110030963 List:
References
50-293-LR, ASLBP 06-848-02-LR, NRC000001, RAS 19379
Download: ML110600910 (130)
v • d • e


Text

Nuclear Regulatory Commission Exhibit # - NRC000001-00-BD01 Docket # - 05000293 Identified: 02/22/2011 Admitted: 02/22/2011 Withdrawn:

Rejected: Stricken:

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Attachment E Severe Accident Mitigation Alternatives Analysis Attachment E contains the following sections.

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

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's 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.1-27 E.1.2.1 Containment Performance Analysis ............................. E.1-27 E.1.2.2 Radionuclide Analysis ........................ E.1-33 E. 1.2.2.1 Introduction .................................... E. 1-33 E.1.2.2.2 Timing of Release ........................ E.1-33 E.1.2.2.3 Magnitude of Release ...................... E.1-34 E.1.2.2.4 Release Category Bin Assignments ....... ................. E.1-34 E.1.2.2.5 Mapping of Level 1 Results into the Various Release Categories . E.1-35 E.1.2.2.6 Collapsed Accident Progression Bins Source Terms .... ....... E.1-43 E.1.2.2.7 Release Magnitude Calculations .......................... E.1-52 E.1.3 IPEEE Analysis .......................... E.1-52 E.1.3.1 Seismic Analysis .......................... E.1-52 E.1.3.2 Fire Analysis ....................... . . . . E.1-52 E.1.3.3 Other External Hazards ...................... E.1-54 E.1.4 PSA Model Peer Review and Difference between Current PSA Model and 1995 Update IPE .. ..................... E.1-54 E.1.4.1 PSA Model Peer Review ........................ .... E.1 -54 E.1.4.2 Major Differences between the Updated IPE PSA Model and 1995 Update IPE Model ................. E.1-55 E.1.4.2.1 Core Damage - Comparison to the PNPS 1995 Update IPE Model ......................... E.1-55 E.1.4.2.2 Containment Performance - Comparison to the Original PNPS IPE Model .......... E.1-59 E.1.5 The MACCS2 Model - Level 3 Analysis . . ............................ E.1-60 E.1.5.1 Introduction ................. I........ E.1-60 E.1.5.2 Input ................ E.1-60 E.1.5.2.1 Projected Total Population by Spatial Element ..... .......... E.1-61 E.1.5.2.2 Land Fraction .................... ........ E.1-62 E.1.5.2.3 Watershed Class ..... E.1-62 i

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report ili ibl Operating License Renewal Stage E.1.5.2.4 Regional Economic Data -.............. ................... .

E.1.5.2.5 Agriculture Data .................... ................... .

E.1.5.2.6 Meteorological Data ................ ................... .

E.1.5.2.7 Emergency Response'Assumptions. E.1-64 .

E.1.5.2.8 Core Inventory ..................... 1-64 .

E.1.5.2.9 Source Terms ........................ 'E.1-66 .

E.1.5.3 Results ................................ E.1-66 .

i E.1.6 References ............................... E.1-69 .

E.2 EVALUATION OF 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 E.2.3 Final Screening and Cost Benefit Evaluation of SAMA Candidates (Phase II) E.2-2 E.2.4 Sensitivity Analyses ............................................ E.2-11 E.2.5 References ................. ......................... E.2-13 ii

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Ij List of Tables Table E.1-1 Core Damage Frequency Uncertainty ....................................... E.1-2 Table E.1-2 PNPS PSA Model CDF Results by Major Initiators ........ E.1-3 Table E.1-3 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs ................ E.1-4 Table E.1-4 Summary of PNPS PSA Core Damage Accident Class ......................... E.1-28 Table E.1-5 Notation and Definitions for PNPS CET Functional Nodes Description ............. E.1-29 Table E.1-7 PNPS Release Categories ..................... E.1-35 Table E.1-6 Release Severity and Timing Classification Scheme Summary ................... E.1-35 Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States ....... E.1-36 Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions ...................... E.1 -44 Table E.1-10 Summary of PNPS Containment Event Tree Quantification ...................... E.1-49 Table E.1-11 Collapsed Accident Progression Bin (CAPB) Source Terms ..................... E.1-50 Table E.1-11 Collapsed Accident Progression Bin (CAPB) Source Terms (continued) ............................................................ E. 1-51 Table E.1-12 PNPS Fire Updated Core Damage Frequency Results ......................... E.1-53 Table E.1-13 Estimated Population Distribution within a 50-mile Radius ....................... E.1-61 iii &:

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-14 PNPS Core Inventory (Becquerels) ........................................ E.1-65 Table E.1-15 Base Case Mean PDR and OECR Values ................................... E.1-67 Table E.1-16 Summary of Offsite Consequence Sensitivity Results .......................... E.1-68 Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation ..... E.2-15 Table E.2-2 Sensitivity Analysis Results..............................................E.245 iv

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage List of Figures Figure E.1-1 PNPS Radionuclide Release Category Summary ......................... E.1-31 Figure E.1-2 PNPS Plant Damage State Contribution to LERF ........... .............. E.1-32 v0

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage ATTACHMENT E.1 EVALUATION OF PSA MODEL

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's 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 the MACCS2 code. The CAFTA code was used to develop the Pilgrim Nuclear Power Station (PNPS) PSA Level I and Level 2 models. This section provides the description of PNPS 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 I Analysis The PSA model (Level I and Level 2) used for the SAMA analysis was the most recent internal events risk model for PNPS (Revision 1, April 2003) [Reference E.1-1]. The PNPS PSA model and documentation has been updated to reflect the current plant operating configuration and design changes as of September 2001. The current PSA model reflects the accumulation of additional plant operating history and component failure and unavailability data as of December 2001. The PSA model also resolves all findings and observations during the industry peer review of the model, conducted in March 2000 [Reference E.1-1]. The PNPS model adopts the small event tree/ large fault tree approach and uses the CAFTA code for quantifying CDF. The Level I and Level 2 PNPS PSA analyses were originally developed and submitted to the NRC in September 1992 as the Pilgrim Nuclear Power Station Individual Plant Examination (IPE)

Submittal [Reference E.1-2].

The PSA model has been updated since the IPE due to the following.

  • In 1995, the original IPE model was changed in response to the NRC Request for Additional Information (RAI) received in April 1995 [Reference E.1-3]. Overall CDF was reduced from 5.85E-5/yr to 2.84E-5/yr. The reduction in CDF was due to removal of HPCI room cooling dependency, revised ADS success criteria, and improved HPCI/RCIC performance.
  • 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.

The PSA model contains the major initiators leading to core damage with baseline CDFs listed in Table E.1-2 [Reference E.1-1].

The current PNPS PSA 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% and E.1-1

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage are judged to be highly unlikely contributors for the identification of cost-beneficial enhancements. These basic events, including component failures, operator actions, and initiating events, were reviewed to determine if additional SAMA actions may need to be considered.

Table E.1-3 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 PNPS PSA model and the SAMAs evaluated in Section E.2.

The uncertainty associated with CDF was estimated using Monte Carlo techniques implemented in CAFTA for the base case mode. The results are shown in Table E.1-1.

Table E.1-1 Core Damage Frequency Uncertainty Confidence CDF (IRY)

Mean value 6.68E-6 5 th percentile.30E-6 5 0 th percentile 5.93E-6 v 95th percentile 1.08E-5 The values in Table E.1-1 reflect the uncertainties associated with the data distributions used in the analysis. The ratio of the 9 5 th percentile to the mean is about 1.62. This uncertainty factor is included in the factor of 6 used to determine the upper bound estimated benefit described in Appendix E, Section 4.21.5.4.

E.1-2

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-2 PNPS PSA Model CDF Results by Major Initiators HE Type IE Description CDF Percentage of TDC Loss of DC Power Buses 3.06E-06 47.77%

LOOP Loss of Offsite Power 1.29E-06 20.12%

TAC Loss of AC Power Buses 8.83E-07 13.78%

LSSW Loss of Salt Service Water 3.91E-07 6.10%

TRAN Transients 3.60E-07 5.62%

LOCA Loss of Coolant Accident 1.75E-07 2.73%

SBO Station Blackout 1.46E-07 2.28%

ATWS Anticipated Transient Without Scram 5.30E-08 0.83%

ISLOCA Interfacing System LOCA 3.64E-08 0.57%

FLOOD Internal Flooding 1.28E-08 0.20%

Total 6.41E-06 100.00%

E.1-3

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR AC" Pilgrim Nuclear Power Station Applicant's Environmental Report OperatinQ License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition IE-T1 6.70E-02 1.337 Loss of offsite This term represents the LOOP initiating event. Industry efforts power (LOOP) over the last twenty years have led to a significant reduction in plant scrams from all causes. Improvements related to enhancing offsite power availability or reliability and coping with SBO events were already implemented and evaluated during Phase I SAMA screening. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.

IE-TDCB 2.63E-03 1.319 Transient caused This term represents an initiating event caused by a complete by loss of 125VDC loss of l25VDC buses D-17, D5, and D37 and random failures of bus B battery D-2. Phase I SAMAs to improve battery charging capability and replace existing batteries with more reliable ones have already been installed. Phase IISAMAs 025,026, 027, 031, 032, 033, 034, and 035 for enhancing DC system availability and reliability were evaluated.

IE-TDCA 2.63E-03 1.304 Transient caused This term represents an initiating event caused by a complete by loss of 125VDC loss of 125VDC buses D-16, D4, and D36, and random failures of bus A battery D-1. Phase I SAMAs to improve battery charging capability and replace existing batteries with more reliable ones have already been installed. Phase IISAMAs 025, 026, 027, 031, 032, 033, 034, and 035 for enhancing DC system availability and reliability were evaluated.

E.1-4

Exhibit No. NRC000001 3J J NRC - Applicant's Environmental Report SAMA Analysis J Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition FXT-XHE-FO-V4T2 2.31 E-02 1.121 Operator fails to This term represents operator failure to align fire water via the align fire water LPCI injection path for alternate RPV vessel injection. Phase I crosstie for reactor SAMAs, including improvement of procedures and installation of pressure vessel instrumentation to enhance the likelihood of success of operator (RPV) injection via action in response to accident conditions, have already been LPCI (transient) implemented. Phase II SAMAs 057 and 059, which recommend proceduralizing use of the diesel fire pump hydroturbine following EDG A failure, and providing a redundant path from fire water pump discharge to LPCI loops A and B cross-tie, were evaluated.

AC2-PHN-PE-23kV 5.OOE-01 1.079 Loss of shutdown This term represents loss of the shutdown transformer 23kV feed transformer 23kV to 4.16kV bus A8. Improvements related to enhancing offsite feed power availability or reliability and coping with SBO events were already implemented and evaluated during Phase I SAMA screening. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.

IE-TSSW 6.85E-05 1.065 Loss of salt service This term represents an initiating event caused by a complete water (SSW) loss of the service water system. Phase I SAMAs were system implemented to improve service water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps. Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.

E.1-5

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding 1001 -

SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition IE-TAC6 2.63E-03 1.059 Transient caused This term represents loss of 4.16kV bus A6. Phase I SAMAs to by loss of 4160VAC improve 4.16kV bus cross-tie capability and revise procedures to bus A6 repair or replace failed 4.16kV breakers have already been implemented. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.

CIV-XHE-FO-DTV 3.01 E-03 1.057 Operator fails to This term represents operator failure to recognize the need to vent containment vent the torus for pressure reduction during loss of containment using direct torus heat removal accident sequences. Phase I SAMAs, including vent (DTV) improvement of procedures and installation of instrumentation to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented. Phase II SAMA 053 to control containment venting within a narrow pressure band to prevent rapid containment depressurization during venting was evaluated.

IE-TAC5 2.63E-03 1.052 Transient caused This term represents an initiating event caused by loss of 4.16kV by loss of 4160VAC bus AS. Phase I SAMAs to improve 4.16kV bus cross-tie bus AS capability and revise procedures to repair or replace failed 4.16kV breakers have already been implemented. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.

E.1-6

9 I NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 J

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition RHR-MAI-MA-HTXAP 4.08E-04 1.051 RHR heat This term represents RHR heat exchanger E-207A unavailable exchanger E-207A due to maintenance, leading to loop A RHR suppression pool unavailable due to cooling and drywell spray modes being unavailable for maintenance containment pressure reduction. Phase I SAMAs have already been implemented to use firewater for drywell spray and to use venting via DTV path to reduce containment pressure. Phase II SAMAs 001, 009, 014, and 059, to provide alternate means of suppression pool cooling and drywell spray and to enhance the availability and reliability of firewater for reactor vessel injection and drywell spray, were evaluated.

RBC-MAI-MA-LOOPA 3.71 E-04 1.046 RBCCW loop A out This term represents RBCCW loop A unavailable due to for maintenance maintenance. A Phase I SAMA was implemented to improve RBCCW system reliability by making component cooling water trains separate. Phase II SAMA 055 to improve RBCCW system reliability by reducing common dependencies was evaluated.

FXT-XHE-FO-DWS 2.21 E-02 1.046 Operator fails to This term represents operator failure to align fire water via the align fire water LPCI injection path for alternate drywell spray to remove cross-tie for drywell containment heat. Phase I SAMAs, including improvement of spray procedures and installation of instrumentation to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented. Phase II SAMAs 057 and 059, which recommend proceduralizing use of the diesel fire pump hydroturbine following EDG A failure, and providing a redundant path from fire water pump discharge to LPCI loops A and B cross-tie, were evaluated.

E.1-7

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition AC8-CBR-CO-204 9.50E-05 1.044 480V circuit breaker This term represents random failure of 480V circuit breaker 52-52-204 fails to 204, leading to loss of power to 480V motor control center (MCC) remain closed B14 and its associated loads. A Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 030 and 058 to improve 480V bus availability were evaluated.

AC8-CBR-CO-103 9.50E-05 1.044 480V circuit breaker This term represents random failure of 480V circuit breaker 52-52-103 fails to 103, leading to loss of power to 480V MCC B15 and its remain closed associated loads. A Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 030 and 058 to improve 480V bus availability were evaluated.

FXT-ENG-FR-P140 1.92E-02 1.043 Diesel fire pump P- This term represents diesel fire pump P-140 failure to run. Phase 140 fails to run 11SAMA 045, to add a diverse injection system and provide an injection source other than fire water, was evaluated.

LCI-HTX-VF-E207A 3.24E-04 1.04 Loop B heat This term represents random failure of RHR heat exchanger E-exchanger E-207A 207A, leading to loop A RHR suppression pool cooling and failure drywell spray modes being unavailable for containment pressure reduction. Phase I SAMAs have already been implemented to use firewater for drywell spray and to use venting via DTV path to reduce containment pressure. Phase II SAMAs 001, 009, 014, and 059, to provide alternate means of suppression pool cooling and drywell spray and to enhance the availability and reliability of firewater for reactor vessel injection and drywell spray, were evaluated.

E.1-8

3 3 NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 Pilgrim LR Proceeding J

50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition LCI-HTX-VF-E207B 3.24E-04 1.039 Loop A heat This term represents random failure of RHR heat exchanger E-exchanger E-207B 207B, leading to loop B RHR suppression pool cooling and failure drywell spray modes being unavailable for containment pressure reduction. Phase I SAMAs have already been implemented to use firewater for drywell spray and to use venting via DTV path to reduce containment pressure. Phase II SAMAs 001, 009, 014, and 059, to provide alternate means of suppression pool cooling and drywell spray and to enhance the availability and reliability of firewater for reactor vessel injection and drywell spray, were evaluated.

IE-T2 8.90E-02 1.038 Loss of PCS This term represents an initiating event caused by a transient with transients PCS unavailable. Industry efforts over the last twenty years have led to a significant reduction of plant scrams from all causes.

Phase II SAMA 038, to improve MSIV design and mitigate the consequences of this event, was evaluated.

RHR-MAI-MA-HTXBP 2.69E-04 1.032 RHR heat This term represents RHR heat exchanger E-207B unavailable exchanger E-207B due to maintenance, leading to loop B RHR suppression pool unavailable due to cooling and drywell spray modes being unavailable for maintenance containment pressure reduction. Phase I SAMAs have already been implemented to use firewater for drywell spray and to use venting via DTV path to reduce containment pressure. Phase II SAMAs 001, 009, 014, and 059, to provide alternate means of suppression pool cooling and drywell spray and to enhance the availability and reliability of firewater for reactor vessel injection and drywell spray, were evaluated.

E.1-9

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition RBC-MAI-MA-LOOPB 2.36E-04 1.029 RBCCW loop B out This term represents RBCCW loop B unavailable due to for maintenance maintenance. A Phase I SAMA was implemented to improve RBCCW system reliability by making component cooling water trains separate. Phase II SAMA 055 to improve RBCCW system reliability by reducing common dependencies was evaluated.

DWS-XHE-FO-W2 2.85E-04 1.026 Operator fails to This term represents operator failure to align the drywell spray align drywell spray mode of RHR for containment pressure reduction. Phase I mode of RHR SAMAs, including improvement of procedures and installation of instrumentation to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented. No additional Phase 1ISAMAs were recommended for this subject.

SPC-XHE-FO-WI 1.54E-04 1.026 Operator fails to This term represents operator failure to align the suppression align suppression pool cooling mode of RHR for containment pressure reduction.

pool cooling mode Phase I SAMAs, including improvement of procedures and of RHR installation of instrumentation to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented. No additional Phase II SAMAs were recommended for this subject.

LCS-CCF-PG-STNRS 2.22E-05 1.024 Common cause This term represents common cause failure of the core spray and failure of strainers RHR suction strainers. A Phase I SAMA, installing improved BS-8002A&B passive emergency core cooling system (ECCS) suction plugged strainers, has been implemented. Phase II SAMAs 042, 044, and 045, which recommend addition of independent injection systems to mitigate this failure event, were evaluated.

E.1-10

3 3 NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 3

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition DC1-CBR-CO-7216A 5.11E-05 1.023 125VDC circuit This term represents random failure of 125VDC circuit breaker breaker 72-16A 72-16A, leading to loss of DC power to bus D-16. Phase I fails to remain SAMAs to improve battery charging capability and replace closed existing batteries with more reliable ones have already been installed. Phase II SAMAs 025, 026, 027, 031, 032, 033, 034, and 035 for enhancing DC system availability and reliability were evaluated.

ADS-XHE-FO-XlT2 6.88E-04 1.023 Operator fails to This term represents operator failure to manually open the SRVs perform emergency for depressurization during transients. Phase I SAMAs, including depressurization improvement of procedures and installation of instrumentation to (transient) enhance the likelihood of success of operator action in response to accident conditions, have already been implemented. No additional Phase II SAMAs were recommended for this subject.

DCl-CBR-CO-72165 5.11E-05 1.023 125VDC circuit This term represents random failure of DC circuit breaker 72-165 breaker 72-165 fails to provide power to DTV valve AO 5025, causing failure of the to remain closed valve to open on demand, resulting in loss of containment venting capability. Phase II SAMA 056 to improve DTV valve availability was evaluated.

OSP-SBO 7.64E-02 1.023 Operator fails to This term represents operator failure to start or align the SBO start or align station diesel to either bus A5 or A6 during a LOOP event. Phase I blackout (SBO) SAMAs, including improvement of SBO procedures and training diesel to either bus to enhance the likelihood of success of operator action in AS or A6 response to accident conditions, have already been implemented.

No additional Phase II SAMAs were recommended for this subject.

E.1-11

Exhibit No. NRC000001 r?1 .11 NRC - Applicant's001-Environmental Report Pilgrim LR Proceeding Ago", I SAMA Analysis

- tll-- I 50-293-LR, 06-848-02-LR TL -

Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition DCI-CBR-CO-7217A 5.11E-05 1.023 125VDC circuit This term represents random failure of 125VDC circuit breaker breaker 72-17A 72-17A, leading to loss of DC power to bus D-17. Phase I fails to remain SAMAs to improve battery charging capability and replace closed existing batteries with more reliable ones have already been installed. Phase II SAMAs 025, 026, 027, 031, 032, 033, 034, and 035 for enhancing DC system availability and reliability were evaluated.

OSP-14 4.10E-02 1.022 Failure to recover This term represents operator failure to recover offsite power offsite power within within 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> during a LOOP event. Phase I SAMAs, including 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> improvement of SBO procedures and training to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented. No additional Phase 11SAMAs were recommended for this subject.

IE-T3A 8.60E-01 1.022 Transients with This term represents an initiating event caused by atransient with condenser initially PCS available. Industry efforts over the last twenty years have available led to a significant reduction of plant scrams from all causes.

Phase II SAMA 038 to improve MSIV design and mitigate the consequences of this event was evaluated.

FXT-MAI-MA-P140 9.22E-03 1.019 Diesel driven fire This term represents diesel fire pump P-140 in maintenance.

water pump P-140 Phase II SAMA 045, to add a diverse injection system and unavailable due to provide an injection source other than fire water, was evaluated.

maintenance E.1-12

Exhibit No. NRC000001 J

NRC - Applicant's Environmental Report SAMA Analysis Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition AC4-RCK-NO-604 2.51 E-03 1.019 4.16kV circuit This term represents failure of the control circuit of 4.16kV circuit breaker 152-604 breaker 152-604, leading to LOOP to safety bus A6. Phase I control circuit no SAMAs to improve 4.16kV bus cross-tie capability and revise output procedure to repair or replace failed 4.16kV breakers have already been installed. In addition, a Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.

DC1-CBR-CO-72175 5.11E-05 1.018 125VDC circuit This term represents random failure of DC circuit breaker 72-175 breaker 72-175 fails to provide power to DTV valve AO 5042B, causing failure of the to remain closed valve to open on demand, resulting in loss of containment venting capability. Phase II SAMA 056 to improve DTV valve availability was evaluated.

CIV-RCK-NO-5042B 2.50E-03 1.018 SV 5042B control This term represents random failure of the control circuit of DTV circuit failure valve AO 5042B, causing failure of the valve to open on demand, resulting in loss of containment venting capability to control containment pressure. Phase IISAMA 056 to improve DTV valve availability was evaluated.

CIV-RCK-NO-A5025 2.50E-03 1.018 AO 5025 control This term represents random failure of the control circuit of DTV circuit failure valve AO 5025, causing failure of the valve to open on demand, resulting in loss of containment venting capability to control containment pressure. Phase II SAMA 056 to improve DTV valve availability was evaluated.

E.1-13

Exhibit No. NRC000001 NRC - Applicant's*10--

Environmental Report Pilgrim LR Proceeding

. A#V1 SAMA Analysis 50-293-LR, 06-848-02-LR

- -""-Vic - IL, z Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition AC4-RCK-NO-504 2.51 E-03 1.017 4.16kV circuit This temi represents failure of the control circuit of 4.16kV circuit breaker 152-504 breaker 152-504, leading to LOOP to safety bus A5. Phase I control circuit no SAMAs to improve 4.16kV bus cross-tie capability and revise output procedures to repair or replace failed 4.16kV breakers have already been installed. Inaddition, a Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.

SSW-MDP-FS-P208D 2.022-03 1.017 SSW pump P-208D This term represents random failure of SSW pump P-208D to fails to start on start. Phase I SAMAs were implemented to improve service demand water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps.

Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.

SSW-CCF-FS-3P208 2.26E-05 1.017 Common cause This term represents common cause failure of 3 service water failure of 3 SSW pumps to start. Phase I SAMAs were implemented to improve pumps to start service water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps. Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.

E.1-14

Exhibit No. NRC000001 J J NRC - Applicant's Environmental Report SAMA Analysis J

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition SSW-MDP-FS-P208E 2.02E-03 1.016 SSW pump P-208E This term represents random failure of SSW pump P-208E to fails to start on start. Phase I SAMAs were implemented to improve service demand water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps.

Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.

IE-S1 3.00E-04 1.015 Medium LOCA This term represents the medium LOCA initiating event. Several Phase I SAMAs have been implemented to provide more reliable or diverse high or low pressure injection systems to mitigate this event. Phase II SAMAs 040, 041, 042, 043, 044, and 054 were evaluated to reduce the CDF contribution from medium LOCA.

LCS-STR-PG-8002A 1.20E-04 1.014 ECCS strainer BS- This term represents failure of core spray and RHR suction 8002A plugged strainer BS-8002A. A Phase I SAMA was implemented to install improved passive ECCS suction strainers. Phase II SAMAs 042, 044, and 045, which recommend addition of independent injection systems to mitigate this failure event, were evaluated.

LCS-STR-PG-8002B 1.20E-04 1.014 ECCS strainer BS- This term represents failure of core spray and RHR suction 8002B plugged strainer BS-8002B. A Phase I SAMA was implemented to install improved passive ECCS suction strainers. Phase II SAMAs 042, 044, and 045, which recommend addition of independent injection systems to mitigate this failure event, were evaluated.

E.1-15

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition ADS-XHE-FO-XISI 7.40E-03 1.013 Operator fails to This term represents operator failure to manually open the SRVs perform emergency for depressurization during medium LOCA. Phase I SAMAs, depressurization including improvement of procedures and installation of during medium instrumentation to enhance the likelihood of success of operator LOCA action in response to accident conditions, have already been implemented. No additional Phase II SAMAs were recommended for this subject.

EDG-ENG-FR-EDGB 6.10E-03 1.013 Emergency diesel This term represents random failure of EDG-B, leading to an SBO generator -B (EDG) event. Phase I SAMAs to improve availability and reliability of the fails to continue to EDGs by creating a cross-tie of EDGs fuel oil supply and run installing a backup SBO diesel generator have already been implemented. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035, for enhancing AC or DC system reliability or to cope with LOOP and SBO events, were evaluated.

AC8-CBR-CO-104 9.50E-05 1.013 480V circuit breaker This term represents random failure of 480V circuit breaker 52-52-104 fails to 104, leading to loss of power to 480V MCC B17 and its remain closed associated loads. A Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 030 and 058 to improve 480V bus availability were evaluated.

HCI-MAI-MA-HCITM 1.62E-02 1.013 HPCI unavailable This term represents HPCI system unavailable due to due to maintenance maintenance. Phase I SAMAs to improve availability and reliability of the HPCI system that have already been implemented include raising backpressure trip setpoints and proceduralizing intermittent operation. Additional improvements were evaluated in Phase II SAMAs 040, 041, 042, 043, 044, and 045.

E.1-16

J I

NRC - Applicant's Environmental Report Exhibit No. NRC000001 Pilgrim LR Proceeding 9

SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition SSW-CCF-FR-3P208 5.59E-06 1.012 Common cause This term represents common cause failure of 3 service water failure of 3 SSW pumps to continue to run Phase I SAMAs were implemented to pumps to run improve service water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps. Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.

AC8-CBR-CO-205 9.50E-05 1.012 480V circuit breaker This term represents random failure of 480V circuit breaker 52-52-205 fails to 205, leading to loss of power to 480V MCC B18 and its remain closed associated loads. A Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 030 and 058 to improve 480V bus availability were evaluated.

IE-T3C 4.40E-02 1.012 Inadvertently This term represents an initiating event caused by inadvertent opened relief valve opening of a relief valve. Improvement of the SRV design and SRV reseat reliability, to reduce the probability and consequences of this initiating event, were evaluated in Phase II SAMAs 046 and 050.

RBC-CCF-CC-4MOVS 1.13E-05 1.012 Common cause This term represents common cause failure of RBCCW heat failure of RBCCW exchanger A & B side MOVs to open. A Phase I SAMA was heat exchanger A & implemented to improve RBCCW system reliability by making B side MOVs (4)to component cooling water trains separate. Phase II SAMA 055 to open improve RBCCW system reliability by reducing common dependencies was evaluated.

E.1-17

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding Ar7 SAMA Analysis Ak =- I-I- 50-293-LR, 06-848-02-LR Ar Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition OSP-24 1.41 E-02 1.011 Failure to recover This term represents operator failure to recover offsite power offsite power within within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during a LOOP event. Phase I SAMAs, including 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> improvement of SBO procedures and training to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented. No additional Phase 11SAMAs were recommended for this subject.

SSW-RCI-FE-3828X 3.OOE-04 1.01 Pressure switch This term represents random failure of SSW pressure switch PS-PS-3828X coil fails 3828X, resulting in loss of SSW system loop A. Phase I SAMAs to energize were implemented to improve service water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps. Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.

EDG-MAI-MA-EDGA 6.41E-03 1.01 EDG-A out for This term represents EDG-A out for maintenance, leading to an maintenance SBO event. Phase I SAMAs to improve availability and reliability of the EDGs by creating a cross-tie of EDGs fuel oil supply and installing a backup SBO diesel generator have already been implemented. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035, for enhancing AC or DC system reliability or to cope with LOOP and SBO events, were evaluated.

E.1-18

D0- 3 NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 9

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition EDG-ENG-FR-EDGA 6.1 OE-03 1.01 EDG-A fails to This term represents random failure of EDG-A, leading to an SBO continue to run event. Phase I SAMAs to improve availability and reliability of the EDGs by creating a cross-tie of EDGs fuel oil supply and installing a backup SBO diesel generator have already been implemented. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035, for enhancing AC or DC system reliability or to cope with LOOP and SBO events, were evaluated.

SSW-MOV-OO-V3805 6.62E-04 1.009 SSW TBCCW A This term represents random failure of SSW MOV MO-3805 to go heat exchanger 90% closed, resulting in loss of SSW to RBCCW loop B. A Phase outlet MOV MO- I SAMA was implemented to improve RBCCW system reliability 3805 fails to go by making component cooling water trains separate. Phase II 90% closed SAMA 055 to improve RBCCW system reliability by reducing common dependencies was evaluated.

SSW-MDP-FS-P208B 2.02E-03 1.009 SSW pump P-208B This term represents random failure of SSW pump P-208B to fails to start on start. Phase I SAMAs were implemented to improve service demand water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps.

Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.

SSW-MDP-FS-P208A 2.02E-03 1.009 SSW pump P-208A This term represents random failure of SSW pump P-208A to fails to start on start. Phase I SAMAs were implemented to improve service demand water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps.

Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.

E.1-19

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding

- - de, SAMA Analysis Jr, 50-293-LR, 06-848-02-LR k-I

-

Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition C 5.80E-06 1.009 Reactor Protection This term represents failure of the RPS. Several Phase I SAMAs System (RPS) to minimize the risks associated with anticipated transient without failure scram (ATWS) scenarios have already been installed. No Phase 11SAMAs were evaluated to further improve reliability of RPS.

However, Phase IISAMA 048 to enhance reliability of the standby liquid control system and improve capability to mitigate the consequences of an ATWS event was evaluated.

AC4-RCK-NO-605 2.51 E-03 1.009 4.16kV circuit This term represents failure of the control circuit of 4.16kV circuit breaker 152-605 breaker 152-605, leading to loss of power to safety bus A6.

control circuit no Phase I SAMAs to improve 4.16kV bus cross-tie capability and output procedures to repair or replace failed 4.16kV breakers have already been installed. In addition, a Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.

RCI-TDP-RS-P206 1.52E-02 1.009 RCIC turbine driven This term represents random failure of the RCIC system. Phase I pump P-206 fails to SAMAs to improve availability and reliability of the RCIC system restart after clear that have already been implemented include raising high level signal backpressure trip setpoints and proceduralizing intermittent operation. Additional improvements were evaluated in Phase II SAMAs 040, 041, 042, 043, 044, and 045.

FXT-RCK-NO-P140 2.50E 1.009 Diesel fire pump P- This term represents diesel fire pump P-140 control circuit failure.

140 control circuit Phase II SAMA 045, to add a diverse injection system and no output provide an injection source other than fire water, was evaluated.

E.1-20

3 J NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 3

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition AC4-RCK-NO-508 2.51 E-03 1.008 4.16kV circuit This term represents failure of the control circuit of 4.16kV circuit breaker 152-508 breaker 152-508, leading to loss of powerto 480V load center B1.

control circuit no Phase I SAMAs to improve 4.16kV bus cross-tie capability and output revise procedures to repair or replace failed 4.16kV breakers have already been implemented. Inaddition, a Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.

AC8-RCK-NO-101 2.50E-03 1.008 480V circuit breaker This term represents random failure of 480V circuit breaker 52-52-101 control 101, leading to loss of power to 480V load center BI and its circuit no output associated loads. A Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 030 and 058 to improve 480V bus availability were evaluated.

EDG-MAI-MA-EDGB 4.09E-03 1.008 EDG-B out for This term represents EDG-B out for maintenance, leading to an maintenance SBO event. Phase I SAMAs to improve availability and reliability of the EDGs by creating a cross-tie of EDGs fuel oil supply and installing a backup SBO diesel generator have already been implemented. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035, for enhancing AC or DC system reliability or to cope with LOOP and SBO events, were evaluated.

E.1-21

Exhibit No. NRC000001 NRC - Applicant's Ir-Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition HCI-TDP-FS-PM205 7.53E-03 1.008 HPCI turbine driven This term represents random failure of the HPCI system. Phase l pump P-205 fails to SAMAs to improve availability and reliability of the HPCI system start on demand that have already been implemented include raising backpressure trip setpoints and proceduralizing intermittent operation. Additional improvements were evaluated in Phase II SAMAs 040, 041, 042, 043, 044, and 045.

RBC-CCF-FS-4PUMP 7.35E-06 1.008 Common cause This term represents common cause failure of four RBCCW failure of four pumps to start. A Phase I SAMA was implemented to improve RBCCW pumps to RBCCW system reliability by making component cooling water start trains separate. Phase II SAMA 055 to improve RBCCW system reliability by reducing common dependencies was evaluated.

AC4-RCK-NO-505 2.51 E-03 1.007 4.16kV circuit This term represents failure of the control circuit of 4.16kV circuit breaker 152-505 breaker 152-505, leading to loss of power supply to safety bus control circuit no A5. Phase I SAMAs to improve 4.16kV bus cross-tie capability output and revise procedures to repair or replace failed 4.16kV breakers have already been installed. Inaddition, a Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.

FXT-XVM-CC-511 5.OOE-04 1.007 Manual valve 10- This term represents random failure of manual valve 10-HO-511 HO-511 fails to to open to provide fire water to LPCI loops A and B. This failure open leads to loss of fire water backup for reactor vessel injection and drywell spray. Phase II SAMA 059 to enhance availability of the fire water system was evaluated.

E.1-22

Exhibit No. NRC000001 3 3 NRC - Applicant's Environmental Report SAMA Analysis Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event DescrIption Disposition FXT-XVM-CC-8156 5.00E-04 1.007 Manual valve 8-1-56 This term represents random failure of manual valve 8-1-56 to fails to open open to provide fire water to LPCI loops A and B. This failure leads to loss of fire water backup for reactor vessel injection and drywell spray. Phase IISAMA 059 to enhance availability of the fire water system was evaluated.

RCI-MAI-MA-RCITM 1.97E-02 1.007 RCIC unavailable This term represents RCIC system unavailable due to due to maintenance maintenance. Phase I SAMAs to improve availability and reliability of the RCIC system that have already been implemented include raising backpressure trip setpoints and proceduralizing intermittent operation. Additional improvements were evaluated in Phase II SAMAs 040, 041, 042, 043, 044, and 045.

CIV-AOV-CC-5042B 1.OOE-03 1.007 AO 5042B fails to This term represents random failure of DTV valve AO 5042B to open on demand open on demand, resulting in loss of containment venting capability to control containment pressure. Phase II SAMAs 001, 009, 014, and 059, to provide alternate means of suppression pool cooling and drywell spray and to enhance the availability and reliability of firewater for reactor vessel injection and drywell spray, were evaluated for containment pressure control.

CIV-AOV-CC-A5025 1.OOE-03 1.007 AO 5025 fails to This term represents random failure of DTV valve AO 5025 to open on demand open on demand, resulting in loss of containment venting capability to control containment pressure. Phase II SAMAs 001, 009, 014, and 059, to provide alternate means of suppression pool cooling and drywell spray and to enhance the availability and reliability of firewater for reactor vessel injection and drywell spray, were evaluated for containment pressure control.

E.1-23

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA

1. Analysis le",

-. -7.--.- 7--1-7-1a- - 50-293-LR, 06-848-02-LR I NK Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition CM 3.30E-01 1.006 RPS mechanical This term represents random failure of the RPS. Several Phase I failure SAMAs to minimize the risks associated ATWS scenarios have already been installed. No Phase II SAMAs were evaluated to further improve reliability of RPS. However, Phase II SAMA 048 to enhance reliability of the standby liquid control system and improve ATWS capability to mitigate the consequences of this event was evaluated.

RBC-MAI-MA-P202E 6.71 E-03 1.006 RBCCW pump This term represents RBCCW pump 202E unavailable due to 202E out for maintenance. A Phase I SAMA was implemented to improve maintenance RBCCW system reliability by making component cooling water trains separate. Phase II SAMA 055 to improve RBCCW system reliability by reducing common dependencies was evaluated.

RBC-MAI-MA-P202F 6.44E-03 1.006 RBCCW pump This term represents RBCCW pump 202F unavailable due to 202F out for maintenance. A Phase I SAMA was implemented to improve maintenance RBCCW system reliability by making component cooling water trains separate. Phase II SAMA 055 to improve RBCCW system reliability by reducing common dependencies was evaluated.

IE-TDC-CCF 3.66E-08 1.006 Common cause This term represents an initiating event caused by a complete failure of 125VDC loss of 125VDC buses D-16 and D-17 or random failure of buses A&B batteries D-1 and D-2. Phase I SAMAs to improve battery charging capability and replace existing batteries with more reliable ones have already been installed. Phase II SAMAs 025, 026, 027, 031, 032, 033, 034, and 035 for enhancing DC system availability and reliability were evaluated.

E.1 -24

Exhibit No. NRC000001

3 I9 NRC - Applicant's Environmental Report SAMA Analysis J Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition SPC-MAI-MA-SPCA 3.01 E-03 1.005 Suppression pool This term represents RHR suppression pool cooling loop A cooling loop A out unavailable due to maintenance. Phase I SAMAs to improve for maintenance availability and reliability of the RHR suppression pool cooling mode that have already been implemented include using drywell spray mode and fire protection cross-tie to provide redundant containment heat removal capability. Additional improvements were evaluated in Phase II SAMAs 001 and 014.

SPC-MAI-MA-SPCB 2.91E-03 1.005 Suppression pool This term represents RHR suppression pool cooling loop B cooling loop B out unavailable due to maintenance. Phase I SAMAs to improve for maintenance availability and reliability of the RHR suppression pool cooling mode that have already been implemented include using drywell spray mode and fire protection cross-tie to provide redundant containment heat removal capability. Additional improvements were evaluated in Phase II SAMAs 001 and 014.

DWS-MAI-MA-DWSA 3.18E-03 1.005 Drywell spray loop This term represents RHR drywell spray loop A unavailable due A out for to maintenance. Phase I SAMAs to improve availability and maintenance reliability of the RHR drywell spray mode that have already been implemented include using suppression pool cooling mode and fire protection cross-tie to provide redundant containment heat removal capability. Additional improvements were evaluated in Phase II SAMA 009.

E.1-25

Exhibit No. NRC000001 Yc-_A

, NRC - Applicant's Environmental Report SAMA Analysis r-i 01*

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition ADS-XHE-FO-XIS2 1.45E-03 1.005 Operator fails to This term represents operator failure to manually open the SRVs perform emergency for depressurization during a small LOCA. Phase I SAMAs, depressurization including improvement of procedures and installation of during small LOCA instrumentation to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented. No additional Phase IISAMAs were recommended for this subject.

E.1-26

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.2 PSA Model - Level 2 Analysis E.1.2.1 Containment Performance Analysis The PNPS Level 2 PSA model used for the SAMA analysis is the most recent internal events risk model, which is an updated version of the model used in the IPE [References E.1-2 and E.1-3].

The Level 2 PSA model used for the SAMA analysis, Revision 1, reflects the PNPS operating configuration and design changes as of September 2001. Specifically, the Level 2 model has been updated to incorporate insights from the independent BWROG peer review.

The PNPS 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 MAAP code [Reference E. 14] 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 I analysis involves the assessment of those scenarios that could lead to core damage. A list of the PDS groups and descriptions from the Level 2 analysis is presented in Table E.1-4.

A full Level 2 model was developed for the IPE and completed at the same time as the Level 1 model. The Level 2 model consists of a single containment event tree (CET) 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.1-5.

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 nntontfil fnr incriv haIth affontc tn theg nohlh- Tha frong icnev role infin*nnrnyimqtcmlv

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report kmr' Operating License Renewal Stage Table E.1-4 Summary of PNPS PSA Core Damage Accident Class PoS S l Point  % of Total Group SipiidDescription Estimate CDF LOCAs Large and small break LOCA with initial or long-term loss 1.16E-7 1.80 of core cooling. Core damage results at low or high reactor pressure. For most PDS, late injection and containment heat removal are available.

TRANS Short and long-term transient events. Core damage 2.43E-7 3.79 results at either low or high reactor pressure. Late injection and containment heat removal are available.

SBO SBO involving a loss of high-pressure injection. Core 1.48E-7 2.31 damage results at either low (stuck-open SRV) or high reactor pressure. All accident mitigating functions are recoverable when AC power is restored.

VSLRUPT Vessel rupture event resulting In LOCA beyond ECCS 4.OOE-9 0.06 capability. All PDS result in core damage at low reactor pressure with late injection available.

ATWS Short-term ATWS that leads to early core damage at high 3.39E-8 0.53 reactor pressure following loss of reactivity control and rapid containment pressurization. Reactor coolant system leakage rates associated with boil-off of coolant through the cycling of SRVs/SV with early core melt subsequent to containment overpressure failure. Late injection and containment heat removal are available.

ISLOCA Large and small break interfacing system LOCA outside 4.00E-9 0.06 containment. Core damage results at low or high reactor pressure with a bypassed containment.

TW Containment decay heat removal systems are not 5.86E-6 91.45 available and coolant recirculation to the torus over pressurizes the containment to failure or venting. The torus is saturated.

Total 6.41 E-06 1.OOE+00 E.1-28

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage ( I Table E.1-5 Notation and Definitions for PNPS CET Functional Nodes Description CET Node CET Functional Node Description Plant Damage State This top event represents the initiators considered in the containment Event (PDSEVNT) performance analysis.

RPV Pressure at This top event identifies the status of the reactor pressure vessel (RPV)

Vessel Failure pressure. RPV@VF is set to success when RPV pressure is low.

(RPV@VF) RPV@VF isset to failure when RPV is high.

In-Vessel Cooling This top event addresses the recovery of coolant injection into the vessel Recovery (IN-REC) after core degradation, but prior to vessel breach. This top event considers the possibility of low-pressure injection systems working once the RPV is depressurized.

Vessel Failure (VF) This top event addresses recovery from core degradation within the vessel and the prevention of vessel head thermal attack. Core melt recovery requires the recovery of core cooling prior to core blocking or relocation of molten debris to the lower plenum and thermal attack of the vessel head.

Early Containment This top event node considers the potential loss of containment integrity at, Failure (CFE) or before, vessel failure. Several phenomena are considered credible mechanisms for early containment failure. They may occur alone or in combination. The phenomena are containment isolation failure; containment bypass; containment overpressure failure at vessel breach; hydrogen deflagration or detonation; fuel-coolant interactions (steam explosions); high pressure melt ejection and subsequent direct containment heating; and drywell steel shell melt-through.

Early Release to This top event node considers the importance of early torus pool scrubbing Torus (EPOOL) in mitigating the magnitude of fission products released from the damaged core. Success implies that fission product transport path subsequent to early containment failure is through the torus water and the torus airspace.

That is, the torus pool is not bypassed. Failure involves a release into the drywell.

Debris Cooled Ex- This top event considers the delivery of water to the drywell, via drywell vessel (DCOOL) sprays, or via injection to the RPV and drainage out an RPV breach onto the drywell floor. Success implies the availability of water and the formation of a coolable debris bed such that concrete attack is precluded. Failure implies that the molten core attacks concrete in the reactor pedestal, that core debris remains hot, and sparing of the concrete decomposition products through the melt releases the less volatile fission products to the containment atmosphere.

E.1-29 J,

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-5 Notation and Definitions for PNPS CET Functional Nodes Description (Continued)

CET Node CET Functional Node Description Late Containment This top event addresses the potential loss of containment integrity in the Failure (CFL) long-term. Late containment failure may result from long-term steam and non-condensable gas generation from the attack of molten core debris on concrete.

Late Release to This top event node considers the importance of late torus pool scrubbing in Torus (LPOOL) mitigating the magnitude of fission products released from the damaged core. Success implies that fission product transport path subsequent to late containment failure is through the torus water and the torus airspace. That is, the torus pool is not bypassed. Failure involves a release into the drywell.

Fission Product This top event addresses fission product releases from the fuel into the Removal (FPR) containment and airborne fission product removal mechanisms within the containment structure to characterize potential magnitude of fission product releases to the environment should the containment fail. Failure implies that most of the fission products from the fuel and containment are ultimately released to the environment without mitigation.

Reactor Building This top event is used to assess the ability of the reactor building to retain (RB) fission products released from containment. Success of top event RB is defined to be a reduction of the containment release magnitude.

E.1-30

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR

. Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage fW- '

Ealy Medun Rlease Eary LowFRlease , 1 mo1°/

.0.52% a '/uly HAh Release

.76%

Lat,eHgh Reease O4.

OO/o a\

Eal/ -No Con~imnat Failure Late MWc Jum Release- 1.73%

2A AN LateL R.lease 70.65%

Figure E.1-1 PNPS Radionuclide Release Category Summary E.1-31

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Transierts Interfaang System LOCAs LOCus Aticipated Transiert wthot 201% 2 t 13YO 0.01%

Sacam 39.82% Vesa Rome 0.01%

1. .

Staficn Blackout 57.03%

Figure E.1-2 PNPS Plant Damage State Contribution to LERF E.1-32

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's 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 I 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 [Reference E.1-4].

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. PNPS characterizes the release timing relative to the time at which the release begins, measured from the time of accident initiation. Two timing categories are used: early (0-24 hours) and late (>24 hours).

Based on MAAP calculations for a spectrum of severe accident sequences, PNPS expects that an Emergency Action Level (as defined by the PNPS Emergency Plan) will be reached within the first half hour after accident initiation. Reaching an Emergency Action Level initiates a formal decision-making process that is designed to provide public protective actions. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of accident initiation, the Level 2 analysis assumed that off-site protective measures would be effective. Therefore, the definitions of the release timing categories are as follows.

  • Early releases are CET end-states involving containment failure prior to or at vessel failure or after vessel failure and occurring within 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> measured from the time of accident initiation and for which minimal offsite protective measures would be accomplished.
  • Late releases are CET end-states involving containment failure greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the time of accident initiation, for which offsite measures are fully effective.

E.1-33

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 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 [Reference E.1-5]. The CsI release 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 four distinct radionuclide release categories or bins according to release magnitude as follows:

(1) High (HI) - 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 Csl [Reference E.1-5].1 (2) Medium (MED) - 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 [Reference E. 1-5].2 (3) Low (LO) - A radionuclide release with the potential for latent health effects. This implies a total integrated release of between 0.001% an'd 1% of the initial core inventory of CsI.

(4) Negligible (NCF) - A radionuclide release that is less than or equal to the containment design base leakage. This implies total integrated release of

<0.001% of the initial core inventory of Csl.

The "total integrated release" as used in the above categories is defined as the integrated release within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after RPV failure. If no RPV failure occurs, then the "total integrated release" is defined as the integrated release within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after accident initiation.

E.1.2.2.4 Release Category Bin Assignments Table E.1-6 summarizes the scheme used to bin sequences with respect to magnitude of release, based on the predicted Csl release fraction and release timing. The combi nation of release magnitude and timing produce seven distinct release categories for source terms. These are the representative release categories presented in Table E. 1-7.

1. Once the Csl source term exceeds 0.1, the source term Is large enough that doses above the early fatality threshold can sometimes occur within a population center a few miles from the site.
2. The reference document indicates that for'Csl release fractions of 1 to 10%, the number of latent fatalities is found to be at least 10% of the latent fatalities for the highest release.

E.1-34

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-6 Release Severity and Timing Classification Scheme Summary Release Severity Release Timing Classification Classification Time of Initial Release from Category Csl %Release Category Accident Initiation High Greater than 10 Early (E) Less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Medium i to 10 Low 0.001 to 1 Late (L) Greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Negligible Less than 0.001 Table E.1-7 PNPS Release Categories Timing of Magnitude of Release Release Low Medium High Early Early/Low Early/Med Early/High NCF Late Late/Low Late/Med Late/High C )

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.

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 E.1-35

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 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 I 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.
  • Accident sequence timing. 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 48 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.1-8 provides a description of the PNPS PDS that are used to summarize the Level 1 results.

"ms Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States PDS Description Point Estimate %fD

%ofCDF PDS-1 Long-term LOCA with loss of high-pressure core makeup O.OOE+00 0.00 from HPCI and RCIC, loss of containment heat removal, and failure to depressurize the primary system for low-pressure core makeup. Core damage results at high primary system pressure. Late injection from low-pressure systems (core spray, LPCI, and firewater) is available, provided primary system depressurization occurs. The containment is vented and intact.

PDS-2 Long-term LOCA with loss of both high-pressure core 1.05E-11 <0.001 makeup (HPCI and RCIC) and containment heat removal.

Core damage results at high primary system pressure.

Because containment venting fails, containment failure occurs long-term. Late injection is available from low-pressure systems (core spray, LPCI, and fire water) provided they survive containment failure.

E.1-36

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)

Point PDS Description Estimate %of CDF PDS-3 Short-term LOCA with loss of high-pressure core makeup, 8.68E-08 1.35 and failure to depressurize the primary system for low-pressure core makeup. Core damage occurs at high primary system pressure. Late injection from core spray, LPCI, and firewater is available, provided primary system depressurization occurs. Containment heat removal is available.

PDS-4 Short-term LOCA with loss of high-pressure core makeup, O.OOE+00 <0.001 loss of containment heat removal, and failure to depressurize the primary system for low-pressure core makeup. Core damage occurs at high primary system pressure. Late injection from core spray, LPCI, and firewater is available, provided primary system depressurization occurs. Unlike PDS-3, containment heat removal is unavailable.

PDS-5 Long-term LOCA with loss of high-pressure core makeup and containment heat removal. Core damage occurs at low primary system. Late injection is available from low-0.OOE+00 0.00 Q.

pressure systems (core spray, LPCI, and fire water). The containment is vented and intact.

PDS-6 Long-term large LOCA. High-pressure core makeup from 0.00E+00 0.00 HPCI and RCIC are unavailable due to the large LOCA.

Because containment venting fails, containment failure occurs long-term. Late injection is available from low-pressure systems (core spray, LPCI, and fire water) provided they survive containment failure. Core damage occurs at low primary system pressure.

PDS-7 Short-term large LOCA with loss of core cooling. Core 1.12E-09 0.08 damage results at low primary system pressure. Late injection from firewater cross tie and containment heat removal are available.

PDS0- Short-term large LOCA with loss of core cooling. Core 4.43E-09 0.07 damage results at low primary system pressure. Late injection from firewater cross tie is available. However, unlike PDS-7, containment heat removal is unavailable.

E.1-37

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)

Point PDS Description Estimate l of CDF PDS-9 Short-term LOCA with loss of high and low-pressure core 3.64E-09 0.06%

cooling. Because the primary system is depressurized, core damage results at low primary system pressure. Late injection from SSW system, containment venting, and containment heat removal are available.

PDS-10 Short-term LOCA with loss of high and low-pressure core O.OOE+00 0.00 cooling. Because the primary system is depressurized, core damage results at low primary system pressure. Late injection from SSW system and containment heat removal are available. However, unlike PDS-9, containment venting is not available.

PDS-11 Short-term LOCA with loss of high and low-pressure core O.OOE+00 0.00 cooling. Core damage results at low primary system pressure. Late injection from SSW system is available.

However, unlike PDS-9, containment venting and containment heat removal are unavailable.

PDS-12 Transient with a loss of long-term decay heat removal. Core 2.37E-08 0.37 damage results at high primary system pressure. Late in-vessel and ex-vessel injection is available. The containment is vented and remains intact at the time of core damage.

PDS-13 Transient with a loss of long-term decay heat removal. Core 3.75E-06 58.5 damage results at high primary system pressure. Late in-vessel and ex-vessel injection is available. Unlike PDS-12 containment venting fails.

PDS-14 Short-term transient with failure to depressurize the primary 1.52E-07 2.37 system. Core damage results at high primary system pressure. Late in-vessel and ex-vessel injection is available. Containment heat removal from RHR is available.

PDS Short-term transient with failure to depressurize the primary 5.07E-08 0.79 system. Core damage results at high primary system pressure. Late in-vessel and ex-vessel injection is available. Containment heat removal from RHR is available. However, containment venting is not available.

E.1-38

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report I

Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)

PDS Description Point  % of CDF PDS-16 Short-term transient with failure to depressurize the primary 4.89E-09 0.08 system. Core damage results at high primary system pressure. Late in-vessel and ex-vessel injection is available. Containment heat removal from RHR is not available, but containment venting is available.

PDS-17 Short-term transient with failure to depressurize the primary 2.53E-09 0.04 system. Core damage results at high primary system pressure. Late in-vessel and ex-vessel injection is available. Neither containment heat removal from RHR nor containment venting is available.

PDS-18 Transient with a loss of long-term decay heat removal. 1.56E-06 24.40 Core damage results at low primary system pressure. Late in-vessel and ex-vessel injection is available. The containment is vented and remains intact at the time of core damage.

PDS-19 Transient with a loss of long-term decay heat removal. 5.24E-07 8.18 Cl Core damage results at low primary system pressure. Late in-vessel and ex-vessel injection is available. Unlike PDS-18 containment venting fails.

PDS-20 Long-term transients with loss of core cooling. Core 6.78E-11 0.001 damage results at low primary system pressure. No late injection, but containment heat removal is available.

PDS-21 Short-term transients (IORV) with loss of core cooling. 8.18E-09 0.13 Core damage results at low primary system pressure. Late injection and containment heat removal are available.

PDS-22 Short-term transients with loss of core cooling. Core 1.08E-09 0.02 damage results at low primary system pressure. Late injection and containment heat removal are available.

However, containment venting is not available.

PDS-231 Short-term transients with loss of core cooling. Core O.OOE+00 0.00 damage results at low primary system pressure. Late injection and containment venting are available, but containment heat removal is not available.

PDS-24 Similar to PDS-23, except that containment venting is not 4.98E-09 0.08 available.

E.1-39

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)

Point PDS Description Estimate C of CDF PDS-25 Short-term transients with loss of core cooling. Core 2.57E-09 0.04 damage results at low primary system pressure. No late injection, but containment heat removal and containment venting are available.

PDS-26 Similar to PDS-25, except that containment venting is not 1.24E-08 0.19 available.

PDS-27 Short-term transients with loss of core cooling. Core 4.40E-11 0.001 damage results at low primary system pressure. Late injection and containment heat removal are not available.

However, containment venting is available PDS-28 Short-term transients with loss of core cooling. Core 1.10E-09 0.02 damage results at low primary system pressure. Late injection, containment heat removal and containment venting are not available.

PDS-29 Long-term SBO involving loss of injection at high primary 1.41 E-07 2.21 system pressure from battery depletion. All accident-mitigating functions are recoverable when AC power is restored.

PDS-30 Short-term SBO sequence involving a loss of high-pressure O.OOE+00 0.00 injection at high primary system pressure from loss of all AC power and DC power or failure of SRVs. All accident-mitigating functions are recoverable when offsite power is restored.

PDS-31 Long-term SBO sequence Involving a loss of high-pressure 2.60E-09 0.04 injection due to one stuck-open safety relief valve or long-term failure of HPCI and RCIC and subsequent failure to depressurize the primary system. Core damage results at low primary system pressure. All accident-rnitigating functions are recoverable when offsite power isrestored.

PDS-32 Short-term SBO sequence involving a loss of high-pressure 4.OOE-09 0.06 injection due to two stuck-open safety relief valves or failure of HPCI and RCIC and one stuck-open safety relief valve.

Core damage results at low primary system pressure. All accident-mitigating functions are recoverable when offsite power is restored.

E.1-40

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.18 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)

PDS Description Point Estimate  % of CDF PDS-33 Short-term large reactor vessel rupture. The resulting loss 4.OOE-09 0.06 of coolant is beyond the makeup capability of ECCS. Core damage occurs in the short term at low primary system pressure. Vessel injection and all forms of containment heat removal (RHR and containment venting) are available.

The containment is not bypassed and AC power is available.

PDS-34 Similar to PDS-33, except that containment heat removal O.OOE+00 0.00 from RHR fails.

PDS-35 Short-term large reactor vessel rupture. The resulting loss O.OOE+00 0.00 of coolant is beyond the makeup capability of ECCS. Core damage occurs in the short term at low primary system pressure. Vessel injection is unavailable. However, all forms of containment heat removal (RHR and containment venting) are available. The containment is not bypassed PDS-36 and AC power is available.

Similar to PDS-35, except that containment heat removal 0.OOE+00 0.00 CW from RHR fails.

PDS-37 Short-term ATWS with failure of SRVs and SVs to open to- 1.95E-08 0.31 reduce primary system pressure. The ensuing primary system over pressurization leads to a LOCA beyond core cooling capabilities. Late injection and containment heat removal are available.

PDS-38 Short-term ATWS that leads to early core damage at low 0.OOE+00 0.00 primary system pressure following successful reactivity control. Late injection is not available. However, containment heat removal is available.

PDS-39 Similar to PDS-38 except that containment heat removal 2.32E-09 0.04 from the RHR system is not available.

PDS-40 Long-term ATWS that leads to late core damage at low 0.OOE+00 0.00 primary system pressure following successful reactivity control. Late injection is available; containment heat removal from the RHR is not available. The containment is vented.

E.1-41 CJ

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)

PDS Description Estinate  % of CDF PDS-41 Short-term ATWS that leads to early core damage at high 1.34E-11 <0.001 primary system pressure following successful reactivity control. Late injection and containment heat removal are available.

PDS-42 Similar to PDS-41 except that containment heat removal 0.00E+00 0.00 from the RHR system Is not available.

PDS-43 Long-term ATWS that leads to late core damage at high 0.OOE+00 0.00 primary system pressure following successful reactivity control. Late injection is available; containment heat removal from the RHR is not available. The containment is vented.

PDS-44 Long-term ATWS that leads to late core damage at high 0.OOE+00 0.00 primary system pressure following successful reactivity control. Late injection is available. However, containment heat removal from the RHR system and containment venting are not available.

PDS-45 Short-term ATWS that leads to containment failure and 3.39E-08 0.53 early core damage at high primary system pressure because of inadequate reactor water level following a loss of reactivity control. Late injection and containment venting are available.

PDS-46 Short-term ATWS that leads to containment failure and 0.OOE+00 0.00 early core damage at high primary system pressure because of inadequate reactor water level following successful reactivity control. No late injection; however, containment venting Is available.

PDS-47 Unisolated LOCA outside containment with early core melt 3.22E-09 0.05 at high RPV pressure.

PDS-48 Unisolated LOCA outside containment with early core melt 7.73E-10 0.01 at low RPV pressure.

E.1-42

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage The PDS designators listed in Table E.1-8 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 then evaluated using a single CET to determine the appropriate release category for each Level 2 sequence. Each end state associated with a Level 2 sequence is assigned to one of the release categories depicted in Table E.1-7. Note, however, that since not all the Level 2 sequences associated with each Level 1 core damage class 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. The CET described in the Level 2 model determines the release category frequency attributed to each Level 1 core damage PDS.

E.1.2.2.6 Collapsed Accident Progression Bins Source Terms The source term analysis results in hundreds of source terms for internal initiators, making calculation with the MACCS2 consequence model cumbersome. Therefore, the source terms were grouped into a much smaller number of source term groups defined in terms of similar properties, with a frequency weighted mean source term for each group.

The consequence analysis source terms groups are represented by collapsed accident progression bins (CAPB). The CAPB were generated by sorting the accident progression bins for each of the forty-eight PDS on attributes of the accident: the occurrence of core damage, the occurrence of vessel breach, primary system pressure at vessel breach, the location of containment failure, the timing of containment failure, and the occurrence of core-concrete 0

interactions. Descriptions of the CAPB are presented in Table E.1-9.

E.1-43

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions CAPB Number Description CAPB-1 [CD, No VB, No CF, No CCI]

Core damage (CD) occurs, but timely recovery of RPV injection prevents vessel breach (No VB). Therefore, containment integrity is not challenged (No CF) and core-concrete interactions are precluded (No CCI). However, the potential exists for in-vessel release to the environment due to containment design leakage.

CAPB-2 [CD, VB, No CF, No CCI]

Core damage (CD) occurs followed by -vessel breach (VB). Containment does not' fail structurally and is not vented (No CF). Ex-vessel releases are recovered, precluding core-concrete interactions (No CCI). Although containment does not fail, vessel breach does occur, therefore the potential exists for in- and ex-vessel releases to the environment due to containment design leakage. RPV pressure is not important because, even though high pressure induced severe accident phenomena (such as direct containment heating [DCH]) occurs, containment does not fail.

CAPB-3 [CD, VB, No CF, CCI]

Core damage (CD) occurs followed by vessel breach (VB). Containment does not fail structurally and is not vented (No CF). However, ex-vessel releases are not recovered in time, and therefore core-concrete interactions occur (CCI). RPV pressure is not important because, even though high pressure induced severe accident phenomena (such as direct containment heating [DCH]) occurs, containment does not fail, nor is the vent limit reached.

CAPB-4 [CD, VB, Early CF, WW, RPV pressure >200 psig at VB, No CCII Core damage (CD) occurs followed by 'vessel breach (VB). Containment fails either II before core damage, during core damage, or at vessel breach (Early CF). i i Containment failure occurs in the torus (WW), above the water level. RPV pressure Is greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena [DCH] are possible). There are no core concrete interactions (No CCI) due to the' presence of an overlying pool of water.

E.1-44

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage C!

Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)

CAPB Description Number CAPB-5 [CD, VB, Early CF, WW, RPV pressure <200 psig at VB, No CCI]

Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).

Containment failure occurs in the torus (WW), above the water level. RPV pressure is less than 200 psig at time of vessel breach; precluding high pressure induced severe accident phenomena. There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.

CAPB-6 [CD, VB, Early CF, WW, RPV pressure >200 psig at VB, CCI]

Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).

Containment failure occurs in the torus (WW), above the water level. RPV pressure is greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena [DCH] are possible). Following containment failure, core-concrete interactions occur (CCI).

CAPB-7 [CD, VB, Early CF, WW, RPV pressure <200 psig at VB, CCI] C)o Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).

Containment failure occurs in the torus (WW), above the water level. RPV pressure is less than 200 psig at time of vessel breach; precluding high pressure induced severe accident phenomena. Following containment failure, core-concrete interactions occur (CCI).

CAPB-8 [CD, VB, Early CF, DW, RPV pressure >200 psig at VB, No CCI]

Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).

Containment failure occurs in the drywell or below the torus water line (DW). RPV pressure is greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena [DCH] are possible). There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.

E.1-45

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)

CAPB Description Number CAPB-9 [CD, VB, Early CF, DW, RPV pressure <200 psig at VB, No CCI]

Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).

Containment failure occurs in the drywell or below the torus water line (DW). RPV pressure isless than 200 psig at time of vessel breach; precluding high pressure induced severe accident phenomena. There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.

CAPB-10 [CD, VB, Early CF, DW,RPV pressure >200 psig at VB, CCI]

Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).

Containment failure occurs in the drywell or below the torus water line (DW). RPV pressure is greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena [OCH] are possible). Following containment failure, core-concrete interactions occur (CCI).

CAPB-11 [CD, VB, Early CF, DW, RPV pressure <200 psig at VB, CCI]

Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).

Containment failure occurs in the drywell or below the torus water line (DW). RPV pressure is less than 200 psig at time of vessel breach; precluding high pressure Induced severe accident phenomena. Following containment failure, core-concrete interactions occur (CCI).

CAPB-12 [CD, VB, Late CF, WW, No CCI]

Core damage JCD) occurs followed by vessel breach (VB). Containment fails late due to loss of containment heat removal (Late CF). Containment failure occurs in the torus (WW), above the water level. RPV pressure is not important because high-pressure severe accident phenomena (such as DCH) did not fail containment.

There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.

E.1-46

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage (I

Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)

CAPB Description Number CAPB-13 [CD, VB, Late CF, WW, CCIX Core damage (CD) occurs followed by vessel breach (VB). Containment fails late (late CF) due to core-concrete interactions (CCI) after vessel breach. Containment failure occurs in the torus (WW), above the water level. RPV pressure is not important because high-pressure severe accident phenomena (such as DCH) did not fail containment.

CAPB-14 [CD, VB, Late CF, DW, No CCI]

Core damage (CD) occurs followed by vessel breach (VB). Containment fails late due to loss of containment heat removal (Late CF). Containment failure occurs in the drywell or below the torus water level (DW). RPV pressure is not important because high-pressure severe accident phenomena did not fail containment. There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.

CAPB-1 5 [CD, VB, Late CF, DW, CCI]

Core damage (CD) occurs followed by vessel breach (VB). Containment fails late (late CF) due to core-concrete interactions (CCI) after vessel breach. Containment failure occurs in the drywell or below the torus water level (DW). RPV pressure is not important because high-pressure severe accident phenomena did not fail containment.

CAPB-16 [CD, VB, BYPASS, RPV pressure >200 psig, No CCI]

Small break interfacing system LOCA outside containment occurs. Core damage (CD) and subsequent vessel breach (VB) results at high RPV pressure with a bypassed containment. There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.

CAPB-17 [CD, VB, BYPASS, RPV pressure <200 psig, No CCI]

Large break interfacing system LOCA outside containment occurs. Core damage (CD) and subsequent vessel breach (VB) results at low RPV pressure with a bypassed containment. There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.

E.1-47

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)

NuCber Description CAPB-18 [CD, VB, BYPASS, RPV pressure >200 psig, CCI]

Small break interfacing system LOCA outside containment occurs. Core damage (CD) and subsequent vessel breach (VB) results at high RPV pressure with a bypassed containment. Following vessel breach, core-concrete interaction occurs (CCI).

CAPB-19 [CD, VB, BYPASS, RPV pressure <200 psig, CCI]

Large break interfacing system LOCA outside containment occurs. Core damage (CD) and subsequent vessel breach (VB) results at low RPV pressure with a bypassed containment. Following vessel breach, core-concrete interaction occurs (CCI).

Il Based on the above binning methodology, the salient Level 2 results are summarized in Tables

%mv E.1-10 and E.1-11 respectively. Table E.1-10 summarizes the results of the CET quantification.

This table identifies the total annual release frequency for each Level 2 release category.

Table E.1-11 provides the frequency, time, duration, energy, and elevation of release for each CAPB.

E.1-48

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage (p-Table E.1-10 Summary of PNPS Containment Event Tree Quantification Release Category Release Frequency (Timing/Magnitude) (/RY)

Late Low 4.53E-06 Late Medium 1.56E-06 Late High O.OOE-00 Early Low 3.32E-08 Early Medium 6.48E-08 Early High 1.13E-07 No Containment Failure 1.11E-07 Nomenclature Timing L (Late) - Greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> E (Early) - Less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> JW)

Magnitude NCF (Little to no release) - Less than 0.001% Csl LO (Low) - 0.001 to 1% Cs1 MED (Medium) -1 to 10% Csl Hi (High) - Greater than 10% Csl E.1-49

("V

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-11 Collapsed Accident Progression Bin (CAPB) Source Terms  !

CAPB Warning Release Release Release Elevation

- CAPB Frequency Time Start I Duration Energy (m)

(Iyear) (sec) (sec) (sec) (W) 1 CAPB-1 9.51 E-08 3.98E+03 3.OOE+01 2.20E+04 9.OOE+03 2.61E+05 2 CAPB-2 1.27E-08 3.96E+03 3.OOE+01 2.20E+04 9.OOE+03 2.50E+05 3 CAPB-3 2.39E-09 3.96E+03 3.OOE+01 2.20E+04 9.OOE+03 2.50E+05 4 CAPB-4 3.29E-09 7.96E+03 3.OOE+01 1.83E+04 3.56E+03 1.IOE+07 5 CAPB-5 2.73E-09 1.31 E+04 3.OOE+01 2.53E+04 7.93E+03 8.34E+06 6 CAPB-6 7.95E-09 1.33E+04 3.OOE+01 2.56E+04 8.11E+03 8.23E+06 7 CAPB-7 7.93E-09 1.38E+04 3.OOE+01 2.61 E+04 8.46E+03 8.03E+06 8 CAPB-8 2.06E-08 9.18E+03 3.00E+01 2.OOE+04 4.59E+03 1.04E+07 9 CAPB-9 9.25E-09 9.21 E+03 3.OOE+01 2.44E+04 8.87E+03 4.18E+06 10 CAPB-10 8.53E-08 1.37E+04 3.OOE+01 2.60E+04 8.40E+03 8.06E+06 11 CAPB-11 4.35E-08 1.37E+04 3.OOE+01 2.60E+04 8.40E+03 8.06E+06 12 CAPB-12 1.70E-06 2.84E+04 3.OOE+01 4.64E+04 9.OOE+03 7.59E+06 13 CAPB-13 2.30E-09 9.14E+03 3.OOE+01 2.71E+04 9.OOE+03 1.80E+06 14 CAPB-14 2.26E-06 2.66E+04 3.OOE+01 4.46E+04 9.OOE+03 7.08E+06 15 CAPB-15 2.12E-06 2.81 E+04 3.OOE+01 4.62E+04 9.OOE+03 7.60E+06 16 CAPB-16 1.18E-09 3.96E+03 3.OOE+01 2.12E+04 9.OOE+03 2.50E+05 17 CAPB-17 6.91E-09 3.96E+03 3.OOE+01 2.14E+04 9.OOE+03 2.50E+05 18 CAPB-18 4.61E-10 3.96E+03 3.OOE+01 2.12E+04 9.OOE+03 2.50E+05 19 CAPB-19 2.43E-08 3.96E+03 3.OOE+01 2.18E+04 9.OOE+03 2.50E+05 E.1-50:

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-11 Collapsed Accident Progression Bin (CAPB) Source Terms (continued)

I Release Fractions NG T Cs Te Sr Ru La Ce Ba 1 1.99E-07 1.85E-07 1.85E-07 O.OOE+O0 1.24E-09 8.OOE-09 5.01E11 8.43E-11 1.70E-08 2 9.97E-05 4.81 E-05 4.66E-05 1.76E-07 3.97E-07 4.OOE-06 1.65E-08 5.15E-08 4.87E-06 3 9.97E-05 5.37E-05 4.97E-05 1.76E-06 5.80E-07 4.OOE-06 2.37E-08 1.57E-07 4,95E-06 4 1.OOE+00 4.90E-02 2.62E-02 4.18E-05 2.46E-05 3.66E-04 8.97E-07 3.04E-06 1.92E-04 5 9.85E-01 7.86E-02 3.68E-02 4.28E-05 4.1OE-05 3.66E-04 1.56E-06 6.79E-06 3.44E-04 6 1.OOE+00 4.02E-02 2.32E-02 1.48E-03 3.19E-04 3.66E-04 6.50E-06 7.17E-05 3.23E-04 7 9.76E-01 6.11 E-02 2.94E-02 1.26E-03 2,30E-04 3.66E-04 9.1 OE-06 1.06E-04 4.52E-04 8 1.OOE+00 2.98E-01 2.72E-01 3.07E-05 9.89E-04 2.23E-02 4.49E-05 6.57E-05 1.1 5E-02 9 5.97E-01 7.61 E-02 7.07E-02 1.41 E-05 9.72E-04 1.09E-02 3.69E-05 7.63E-05 1.02E-02 10 1.OOE+00 2.80E-01 2.49E-01 1.1 E-02 3.07E-03 1.81E-02 7.95E-05 5.81 E-04 1.03E-02 (-j 11 9.79E-01 1.73E-01 1.41 E-01 9.97E-03 3.13E-03 1.78E-02 1.22E-04 9.39E-04 1.72E-02 12 2.01 E-01 5.84E-05 4.37E-05 1.25E-07 2.36E-07 1.72E-06 8.04E-09 2.56E-08 2.99E-06 13 9.97E-01 7.99E-03 5.99E-03 1.76E-04 3.63E-05 3.66E-04 2.15E-06 1.41 E-05 4.52E-04 14 7.75E-01 2.88E-02 2.67E-02 2.47E-05 2.05E-04 2.13E-03 8.49E-06 2.27E-05 2.61 E-03 15 9.97E-01 2.76E-01 2.68E-41 1.27E-03 2.27E-03 2.25E-02 9.33E-05 3.OOE-04 2.74E-02 16 1.OOE+00 6.71 E-02 3.26E-02 4.06E-04 9.11 E-05 2.21 E-02 1.45E-06 1.65E-05 4.27E-05 17 9.72E-01 3.62E-01 3.37E-01 1.34E-03 2.37E-03 2.20E-02 9.90E-05 1.62E-04 8.57E-03 18 1.OOE+00 9.76E-02 6.25E-02 2.09E-02 4.67E-03 2.27E-02 7.45E-05 8.50E-04 2.12E-03 19 9.72E-01 4.03E-41 3.77E-01 6.87E-02 9.58E-03 2.26E-02 3.OOE-04 2.33E-03 1.20E-02 E.1-51 Q-

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.2.2.7 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 versus time (time of containment failure is determined by comparing these values with the nominal containment capability);
  • radionuclide release time and magnitude for a large number of radioisotopes; and
  • release fractions for twelve radionuclide species.

E.1.3 IPEEE Analysis E.1.3.1 Seismic Analysis PNPS performed a seismic PRA following the guidance of NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (lPEEE) for Severe Accident Vulnerabilities, June 1991. The seismic PRA model was performed in conjunction with the SQUG program in 1994 as part of the IPEEE submittal report [Reference E.1-6]. The seismic, high wind, and external flooding analyses determined that the plant is adequately designed to protect against the effects of these natural events.

A number of plant improvements were identified in Table 2.4 of NUREG-1 742, Perspectives Gained from the IPEEE Program, Final Report, April 2002 [Reference E.1 -8]. These improvements were implemented.

The seismic CDF in the IPEEE was conservatively estimated to be 5.82x10-5 per reactor-year.

The seismic CDF has recently been re-evaluated to reflect the updated Gothic computer code room heat up calculations that predict no room cooling requirements for HPCI, RCIC, Core Spray, and RHR areas; to update random component failure probabilities; and to model replacement of certain relays with a seismically rugged model. The updated seismic CDF of 3.22x10-5 per reactor-year was used in estimation of the factor of 6 used to determine the upper bound estimated benefit described in Section 4.21.5.4.

E.1.3.2 Fire Analysis The PNPS internal fire risk model was performed in 1994 as part of the IPEEE submittal report

[Reference E.1-6]. The PNPS fire analysis was performed using the conservative EPRI's Fire Induced Vulnerability Evaluation (FIVE) methodology for qualitative and quantitative screening of fire areas and for fire analysis of areas that did not screen [Reference E.1 -71. The FIVE methodology is primarily a screening approach used to identify plant vulnerabilities due to fire initiating events.

E.1-52

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 0

Table E.1-12 presents the results of the PNPS IPEEE fire analysis. The values presented in Table E.1-12 are taken from NUREG-1742 [Reference E.1-8]. These values are the same as the original IPEEE fire CDF results (2.20E-5 per reactor-year) [Reference E.1-6] after the response to NRC questions/issues regarding fire-modeling progression. A revised fire zone CDF of 1.91 E-5 per reactor-year, generated to reflect updated equipment failure probability and unavailability values was used in estimation of the factor of 6 used to determine the upper bound estimated benefit described in Section 4.21.5.4.

The significant fire scenarios involve fires occurring in the train B switchgear room, turbine building heater bay, vital motor generator set room, and train A switchgear room.

Table E.1-12 PNPS Fire Updated Core Damage Frequency Results Fire New Compartment Description CDF/year Estimate Sub-Area CDF/year 1E Reactor Building West, El. 21 9.7E-07 8.25E-07 2B Turbine Building Heater Bay 2.1 E-06 2.74E-06 3A Train B RBCCW/TBCCW Pump and Heat 2.0E-06 1.31 E-06 I

Exchanger Room 4A Train A RBCCW[TBCCW Pump and Heat 9.8E-07 2.95E-07 Exchanger Room 6 Control Room 1.6E-06 8.90E-07 7 Cable Spreading Room 9.5E-07 7.85E-07 9 Vital Motor Generator Set Room 2.4E-06 2.38E-06 12 Train A Switchgear Room 3.1E-06 2.30E-06 13 Train B Switchgear Room 6.1E-06 6.85E-06 26 Main Transformer 1.5E-06 7.60E-07 2.2E-05 1.91 E-05 E.1-53

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.3.3 Other External Hazards The PNPS IPEEE submittal [Reference E.1-6], in addition to the internal fires and seismic events, examined a number of other external hazards:

  • external flooding; and
  • ice, hazardous chemical, transportation, and nearby facility incidents.

In consequence of the above external hazards evaluation, no plant modifications were required for PNPS.

No risks to the plant occasioned by high winds and tornadoes, external floods, Ice, and hazardous chemical, transportation, and nearby facility incidents were identified that might lead to core damage with a predicted frequency in excess of 1046/year. Therefore, these other external event hazards are not included in this attachment and are expected not to impact the conclusions of this SAMA evaluation.

E.1.4 PSA Model Peer Review and Difference between Current PSA Model and 1995 Update IPE E.1.4.1 PSA Model Peer Review The original IPE PSA model was peer reviewed on March 2000 using the BWROG PSA Peer Review Certification Implementation Guidelines. Facts and Observation sheets documented the certification teams' insights and potential level of significance. As part of the update of the IPE PSA models, all major issues and observations from the BWROG Peer Review (i.e., Level A, B, C, and D observations) have been addressed and incorporated into the current IPE PSA model, April 2003 [Reference E.1-1].

For the current IPE/PSA model update, individual work packages (event tree, fault tree, human reliability analysis (HRA), data, etc.) and internal flooding analysis were circulated to each PSA member for independent peer review. The accident sequence packages, system work packages, HRA, and internal flooding analyses were also assigned to the appropriate PNPS plant personnel for review. For example, event trees, system analyses, and fault tree models were forwarded to the applicable plant systems engineers and the HRA was assigned to individuals from the plant Operations Training department for review. Similarly, the accident sequence packages, system work packages, HRA report, containment performance analysis, fault tree and event tree models, and Level 2 models were peer reviewed by an outside consultant.

The Entergy license renewal project team and plant staff reviewed consequence and risk estimates for the SAMA analyses.

E.1-54

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage The peer review process emphasized the role of plant staff, external consultants, and BWROG PSA certification in this recent model update. The peer reviews served to ensure the accuracy of both the assumptions made in the models and the results. The results of the peer review and resolutions are presented in Section 5 and Appendix P of the Pilgrim Nuclear Power Station Individual Plant Examination for Internal Events update report, April 2003 [Reference E.1-1].

E.1.4.2 Major Differences between the Updated IPE PSA Model and 1995 Update IPE Model E.1.4.2.1 Core Damage - Comparison to the PNPS 1995 Update IPE Model The current PNPS IPE/PSA update model was completely revised in response to the BWROG Peer Review of March 2000 [Reference E.1-1]. The updated model is based upon all procedures and plant design as of September 30, 2001, and plant data as of December 31, 2001. The results yield a measurably lower CDF (point estimate CDF - 6.41 E-6/reactor year) than the original IPE (point estimate CDF - 5.85E-5/yr) [Reference E.1-2] and 1995 PSA model update (point estimate CDF - 2.84E-5/yr) [Reference E.1-31. (The 1995 update was performed to answer NRC questions following the IPE submittal.) The improved results are due to improved plant performance, replacement of switchyard -breakers, more realistic success criteria based on MAAP runs, and more sophisticated data handling. Major changes are summarized as follows.

A. Initiating Event The initiating event frequencies were updated to include current plant data and recent NRC publication information. For example, the LOOP frequency decreased significantly from the original IPE frequency of 0.475/yr to the current value of 0.067/yr [Reference E.1-1], which reflects the decreased occurrence of LOOP events since 1990 and replacement of switchyard breakers. In addition, fault tree models were developed to calculate support system initiating event frequencies.

B. Accident Sequence Evaluation Event trees from the original IPE were completely revised. BWROG certification findings and observations were incorporated into the revised event trees. Major facts and observations include the following.

(1) LOOP Event Tree The LOOP event was completely revised to account for failure modes of HPCI/RCIC beyond 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of operation; RPV depressurization on HCTL; and transfer to the SBO tree to address such items as premature battery depletion and AC recovery at 30 minutes and beyond.

E.1-55 C>

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage (2) SBO Event Tree Current update reflects GE load shed calculations and use of plant SBO procedures for DC load shedding.

(3) Inadvertent Stuck Open Relief Valve (IORV) Event Tree The IORV event tree was modified to include RPV depressurization with two SRVs given high-pressure injection failure.

(4) LOCAs Event Trees The update considers both HPCI and RCIC for small break LOCAs.

Large and medium LOCAs and subsequent ATWS are modeled as core damage end states in the updated model. Small break LOCAs and ATWS are treated as similar to transient-induced ATWS.

The vapor suppression system is considered during large LOCAs events.

(5) ATWS Event Tree The revised ATWS tree reflects the potential for MSIV closure on low RPV level.

The revised ATWS model takes into consideration "inhibit ADS" and MSIV bypass issues. In addition, HRA values take into consideration ATWS accident progressions for RPV and containment conditions predicted by MAAR (6) Loss-of-Containment Heat Removal Sequences The revised event trees model the potential impact from containment venting on low-pressure system operation. For example, no credit is given for core spray and LPCI if containment venting is required. In addition, other containment related phenomena, such as high torus temperatures (HPCI) and high containment pressures (RCIC, SRVs) are reflected in the updated event trees.

The update model only considers the DTV path for containment venting.

(7) ISLOCA Event Tree NSAC-154 [Reference E.1-10] and NUREG/CR-5124 [Reference E.1-11] were used to reassess the ISLOCA analysis.

Success criteria for low-pressure injection during an ISLOCA are consistent with those used for small LOCAs.

E.1-56

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage The revised ISLOCA event tree credits use of condensate or fire water for large ISLOCA events provided that LPCI or core spray operation had previously occurred to provide initial RPV reflood.

(8) Other Changes The revised event trees credit use of feedwater when appropriate.

Control Rod Drive system flow into the RPV is credited for sequences that involve loss of containment heat removal and subsequent requirement to control containment pressure with direct torus containment venting.

Consistent success criteria were employed for RPV depressurization for transients, medium LOCAs, and small LOCAs.

The revised PNPS IPE models are based on the BWROG EPGs/SAGs Revision 4 of the BWROG EPGs [Reference E.1-1].

Core damage definition has been revised to be consistent with the EPRI PSA Applications Guide [Reference E.1-12]. That is, core damage occurs when peak clad temperature exceeds 22000F.

HPCI and RCIC use is based on a 24-hour mission time. C C. Thermal - Hydraulic (T-H) Analysis T-H analysis has been completely revised and improved to better support the success criteria.

The MAAP4 computer code [Reference E.1-4] was used to update and address the many issues raised by the BWROG certification team, such as the following.

  • A basis was provided for the timing and discharge pressure (flow) adequacy when using the fire water system for successful mitigation during transients and small LOCAs.
  • Success criteria for SORV are same as for non-SORV cases (2 SRVs are required for successful RPV depressurization).
  • Plant specific calculations were performed to identify the plant response for single or double recirculation pump trip failures.
  • The appropriateness of the core damage definition used in the update was verified.

E.1-57

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Hi'" Operating License Renewal Stage In addition to the MAAP4 code, the GOTHIC code [Reference E.1-13] was used to predict various room heatup rates for the reactor building, turbine building, switchgear room, and battery room.

D. System Analysis System fault tree models from the original IPE were completely revised to reflect the as-built plant configuration. MAAP analyses were clearly identified to support the success criteria of these Level 1 models. More detailed modeling for the logic interlock was included in the system models. A detailed fault tree for the RPS was developed based on NUREG/CR-5500 [Reference E.1-9], which decreased the failure-to-scram probability from 3.OE-5/yr to 5.8E-6/yr.

E DataAnalysis Component failure data, both generic and plant-specific, were reviewed and updated with more recent experience (the performance of risk significant systems HPCI and RCIC has greatly improved since the original IPE). Plant-specific data were adjusted for industry experience using Bayesian updates. Maintenance unavailability values were updated based on maintenance rule records from the system engineers. More recent common cause failure data and approach NUREG/CR-5497 [Reference E.1-14] were factored into this update. In particular, a more detailed and refined common-cause failure methodology (Alpha model) has been applied in this update. In addition, more common-cause equipment failure groups such as fans, dampers,

/4gw transformers, DC power panels, and circuit breakers have been included in the analysis.

F. HA A complete revision of the HRA was performed to identify, quantify, and document the pre-initiator and post-initiator human errors (including recoveries). The updated HRA was performed using NUREG/CR-1278 [Reference E.1-15], also referred to as THERP. Screening values were only used for low-significance human errors. In addition, a detailed analysis was performed to treat dependencies between post-initiator errors.

G Dependencv Analysis A complete revision of the internal flooding analysis was developed to systematically address spatial dependencies.

Dependency between pre-initiator human errors (such as miscalibration of instruments) was modeled. In addition, dependencies between multiple post-accident operator actions appearing in the same accident sequence were evaluated. -

Detailed component dependency tables were developed to address the support systems associated with the modeled systems and components.

E.1-58

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage H. Structural Response The ISLOCA frequency was revised.

RPV overpressure and capability of the reactor building were included in the Level 2 assessment.

1. Quantification The truncation value was lowered to I.OE-11.

Human Error Probability (HEP) dependencies and recovery actions in the cutsets were evaluated.

ATWS contribution decreased due to lower probability of failure to scram based on NUREG/CR-5500 [Reference E.1-9].

The HRA was completely revised to address a comment from the PSA Certification [Reference E.1-16] that many of the HEPs were not realistic using the previous methodology. In many cases (e.g., failure to perform DTV), the previous HEPs were judged to be overly conservative.

J. Internal Flooding Analysis The internal flooding analysis from the original IPE was completely revised to include a detailed, systematic examination of the flood source and progression for each of the analyzed flooding scenarios. In addition, the updated internal flooding analysis considers the effects of spray on equipment.

K. Uncertainty Analysis An uncertainty analysis was performed for this update.

E.1.4.2.2 Containment Performance - Comparison to the Original PNPS IPE Model Containment performance analysis models were completely revised from the original IPE.

Propagation of Level 1 cutsets to the Level 2 CET was developed. A detailed LERF model was developed to ensure that LERF calculations are consistent with the PSA Applications Guide and NRC requirements for RG 1.174 [Reference E.1-17]. Other salient items incorporated are the following.

  • CET fault models were revised to ensure that mitigating systems were not degraded in the Level I sequence.
  • CET fault tree models allowed credit for AC power recovery post core damage. This ensures that the models do not allow SBO core damage sequences to benefit from AC supported equipment in Level 2 without explicit consideration of AC power recovery.

E.1-59

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage

  • Shell melt-through phenomena were considered where applicable.
  • Operator responses to key actions were reassessed to incorporate the probability for success given the containment conditions and Emergency Operating Procedure directions.
  • Direct torus venting was considered post core damage.
  • PNPS-specific primary containment structural evaluation was included in the CET. This also included a structural evaluation of torus failure due to dynamic loading during ATWS scenarios, torus break below the water line, and bellows seal capability.
  • A reactor building bypass fault tree model was developed to assess the impact on the Level 2 analysis.

E.1.5 The MACCS2 Model - Level 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 the MACCS2 [Reference E.1-18] was created for PNPS. 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 ninput parameters used to obtain the off-site dose and economic impacts for cost-benefit analyses, E.1.60

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.5.2.1 Projected Total Population by Spatial Element The total population within a 50-mile radius of PNPS was estimated for the year 2032, the end of the proposed license renewal period, for each spatial element by combining total resident population projections with transient population data obtained from Massachusetts and Rhode Island. Table E.1-13 shows the estimated population distribution.

Table E.1-13 Estimated Population Distribution within a 50-mile Radius 0-10 10-20 20-30 30-40 40-50 50-Mile Sector Miles Miles Miles Miles Miles Total N 0 0 0 0 80474 80474 NNE 3 0 0 0 0 3 NE 3 0 0 0 0 3 ENE 3 0 33121 0 0 33124 E 5 0 33121 23185 0 56311 ESE 23 0 49682 92740 0 142445 SE 950 9936 115925 23185 0 SSE 13289 69555 82803 0 0 149996 165647 Q.,

S 23695 99364 132485 84383 43397 383324 SSW 23695 49762 23696 23185 21699 142037 SW 23695 71088 277374 349491 114546 836194 WSW 23695 71088 277374 349491, 183037 904685 W 22818 71088 277374 388324 286370 1045974 WNW 16494 71088 118481 303450 390150 899663 NW 11269 71088 195075 1529212 405561 2212205 NNW 5599 35544 43350 31295 321894 437682 Total 165236 619601 1659861 3197941 1847128 7489767 The 2000 U.S. Census Bureau data, togetherwith Massachusetts and Rhode Island population projection data, was used to project county-level resident populations to the year 2032.

Seasonal peak transient population was conservatively used to establish a transient/resident population ratio for each county within the 50-mile radius. The ratio was found to be decreasing over time. For purposes of this study, the total county level population values were estimated by E.1-61 (C-

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage summing the year 2000 peak transient population of each county and the projected year 2032 permanent population of that county to obtain the 2032 total county population.

E.1.5.2.2 Land Fraction The land fraction for each spatial element was estimated from the PNPS Emergency Planning Zone maps for radii of 2, 5, and 50 miles [Reference E.1-20].

E.1.5.2.3 Watershed Class There are two watershed types in the 50-mile zone surrounding PNPS: ocean and land (watersheds) drained by rivers. There are no major lakes. The watershed index assigns "`" to any spatial element having a non-zero land fraction and "2" to all elements over the Atlantic Ocean or its bays.

E.1.5.2.4 Regional Economic Data RegaLon Index Each spatial element was assigned to an economic region, defined in this report as a county. Where a spatial element covers portions of more than one county, it was assigned to that county having the most area within the element.

Regional Economic Data County level economic data were obtained from the U.S. Department of Agriculture.

The Census of Agriculture is conducted every five years and data from 1997 and 1992 were used to project the farm-related economic data for 2002.

VALWF - Value of Fain, Wealth MACCS2 requires an average value of farm wealth (dollars/hectare) for the 50-mile radius area around PNPS. The county-level farmland property value was used as a basis for deriving this value. VALWF is $23,578/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 $189,041/

person.

Other economic parameters and their values are shown below. The values were obtained by adjusting the economic data from a past census given as default values in Reference E.1-18 with the consumer price index of 177.1, which is the average value for the year 2001, as appropriate.

E.1-62

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report

Operating License Renewal Stage Variable Description Value EVACST Daily cost for a person who has been evacuated 42.3

($/person-day)

POPCST Population relocation cost ($/person) 7840 RELCST Daily cost for a person who is relocated ($/person-day) 42.3 CDFRMO Cost of farm decontamination for the various levels of 881 decontamination ($/hectare) 1959 CDNFRM Cost of non-farm decontamination for the various levels of 4700 decontamination ($/person) 12540 DLBCST Average cost of decontamination labor ($/person-year) 54800 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 New England Agricultural Statistics, 2001. The (. !

crops listed for each of the two states, Massachusetts and Rhode Island, were 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 PNPS site meteorological monitoring system and the Automated Surface Observatory System (ASOS) at Plymouth Airport.

Site Specific Data Site specific meteorological data is available from two meteorological towers, one located off the main parking lot and 'the second located west of the old l&S building, the "lower" and "upper' towers respectively. The upper tower is the designated data source for MACCS2 input. Data from the lower tower was'used only if measurements from the upper tower were missing for a specific hour.

Year 2001 hourly data from the upper tower was used in this analysis. The data was more than 98% complete. Missing data was obtained either from the lower tower or from estimates based on adjacent valid measurements of the missing hour.

E.1 Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Accumulated Precipitation The nearest source of hourly precipitation data to PNPS is the ASOS at Plymouth Airport. The data was converted to MACCS2 input format to provide precipitation in hundredths of an inch.

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. PNPS mixing height data, given in Reference E.1-19, was used for MACCS2 analysis.

E.1.5.2.7 Emergency Response Assumptions Details of the evacuation time estimates including supporting assumptions regarding population, alarm criteria, delay times, areas, speed, distance, and routes are contained in the PNPS Emergency Plan [Reference E.1-20].

Evacuation Delay Time The elapsed time between siren alert and the beginning of evacuation is 40 minutes. A sensitivity case that assumes 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for evacuees to begin evacuation was considered in this study to evaluate consequence sensitivities due to uncertainties in delay time.

Evacuation Speed The worst case for PNPS evacuation is during the winter, under adverse weather conditions, since snow removal can add up to an hour and a half to the evacuation time.

The radius of the Emergency Planning Zone is 10 miles. Assuming that the net movement of the entire population is 10 miles, the time required for evacuation ranges from 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 35 minutes to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 30 minutes, and the average evacuation speed ranges from 2.79 miles/hour in clear weather to 1.54 miles/hour under adverse weather conditions. The average evacuation speed is 2.17 miles/hour, or 0.97 meter/second.

A sensitivity case that assumes a lower evacuation speed of 0.69 meter/second was considered in this study to evaluate consequence sensitivities due to uncertainties in evacuation speed.

E.1.5.2.8 Core Inventory The estimated PNPS core inventory (Table E.1-14) used in the MACCS2 input is based on a power level of 2028 MW(t).

E.1-64

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-14 PNPS Core Inventory (Becquerels)

Nuclide Inventory Nuclide Inventory Co-58 1.15E+16 Te-131m 2.87E+17 Co-60 1.37E+16 Te-132 2.80E+18 Kr-85 1.88E+16 1-131 1.94E+18 Kr-85m 6.84E+17 1-132 2.85E+18 Kr-87 1.24E+18 1-133 4.07E+18 Kr-88 1.68E+18 1-134 4.45E+18 Rb-86 1.05E+15 1-135 3.83E+18 Sr-89 2.08E+18 Xe-133 4.07E+18 Sr-90 1.47E+17 Xe-135 9.68E+17 Sr-91 2.71E+18 Cs-134 3.17E+17 Sr-92 2.83E+18 Cs-136 8.51E+16 Y-90 1.58E+17 Cs-137 1.90E+17 Y-91 2.54E+18 Ba-139 3.75E+18 Y-92 2.84E+18 Ba-140 3.70E+18 Y-93 3.23E+18 La-140 3.77E+18 Zr-95 3.34E+18 La-141 3.48E+18 Zr-97 Nb-95 3.44E+18 3.16E+18 La-142 Ce-141 3.35E+18 3.36E+18 U

Mo-99 3.65E+18 Ce-143 3.27E+18 Tc-99m 3.15E+18 Ce-144 2.18E+18 Ru-103 2.77E+18 Pr-143 3.20E+18 Ru-105 1.85E+18 Nd-147 1.43E+18 Ru-106 7.52E+17 Np-239 4.26E+19 Rh-105- 1.38E+18 Pu-238 2.96E+15 Sb-127 1.74E+17 Pu-239 7.51E+14 Sb-129 6.06E+17 Pu-240 9.41 E+14 Te-127 1.69E+17 Pu-241 1.62E+17 Te-127m 2.27E+16 Am-241 1.65E+14 Te-129 5.68E+17 Cm-242 4.35E+16 Te-129m 1.49E+17 Cm-244 2.35E+15 Source: derived from Reference E.1-21 for a power level of 2028 MW(t)

E.1-65

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.5.2.9 Source Terms Twelve release categories, corresponding to internal event sequences, were part of the MACCS2 input. Details of the source terms for postulated internal events are available in on-site documentation. 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. The base case assumes 40 minute delay and 0.97 meter/sec speed of evacuation. Sensitivity case I is the base case with delayed evacuation of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Sensitivity case 2 is the base case with an evacuation speed of 0.69 meter/sec.

Table E.1-15 shows estimated base case mean risk values for each release mode. The estimated mean values of PDR and offsite OECR for PNPS are 13.6 person-rem/yr and

$45,900/yr, respectively.

E.1-66

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-15 Base Case Mean PDR and OECR Values Pplto Offsite Offsite Release Frequency Popsin Economic Population Dose Economic Cost Modoser CstRisk (PDR) Rik(E )

(person-sv) Cs (person-rem/yr) (S/yr)

CAPB-1 9.51 E-08 4.66E-01 3.82E+06 4.43E-06 2 3.63E-01 CAPB-2 1.27E-08 9.96E+01 6.40E+06 1.26E-04 8.1OE-02 CAPB-3 2.39E-09 1.06E+02 6.48E+06 2.53E-05 1.55E-02 CAPB-4 3.29E-09 1.38E+04 4.28E+09 4.54E-03 1.41E+01 CAPB-5 2.73E-09 1.81E+04 5.30E+09 4.94E-03 1.45E+01 CAPB-6 7.95E-09 1.51 E+04 3.51 E+09 1.20E-02 2.79E+01 CAPB-7 7.93E-09 1.67E+04 4.42E+09 1.32E-02 3.51 E+01 CAPB-8 2.06E-08 4.1OE+04 1.47E+10 8.44E-02 3.03E+02 CAPB-9 9.25E-09 2.37E+04 8.33E+09 2.19E-02 7.70E+01 CAPB-10 8.53E-08 4.31 E+04 1.54E+10 3.68E-01 1.31 E+03 (w

CAPB-11 4.35E-08 3.45E+04 1.15E+10 1.50E-01 5.OOE+02 CAPB-12 1.70E-06 9.72E+01 4.63E+06 1.65E-02 7.88E+OO CAPB-13 2.30E-09 7.30E+03 6.53E+08 1.68E-03 1.50E+OO CAPB-14 2.26E-06 1.58E+04 4.14E+09 3.57E+00 9.36E+03 CAPB-15 2.12E-06 4.31 E+04 1.59E+10 9.14E+00 3.37E+04 CAPB-16 1.18E-09 1.86E+04 5.50E+09 2.19E-03 6.48E+OO CAPB-17 6.91E-09 4.81E+04 1.71E+10 3.32E-02 1.18E+02 CAPB-18 4.61E-10 2.38E+04 7.86E+09 1.1OE-03 3.62E+OO CAPB-19 2.43E-08 5.31 E+04 1.88E+10 1.29E-01 4.56E+02 Totals 1.36E+01 4.59E+04

1. 1 sv= 100 rem
2. 4.43E-06 (person-rem/yr) = 9.51 E-08 (/yr) x 4.66E-01 (person-sv) x 100 (remlsv)

E.1-67

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Results of sensitivity analyses indicate that a delayed evacuation or a lower evacuation speed would not have significant effects on the offsite consequences or risks determined in this study.

Table E.1-16 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 that the maximal deviation is less than 2% between the base case population dose and the Sensitivity Case 2 population dose for release mode CAPB-8.

Table E.1-16 Summary of Offsite Consequence Sensitivity Results Population Dose (person-sv) Offsite Economic Cost ($)

Release 2-Hr Lower 2-Hr Lower Base Case Delayed Speed of Base Case Delayed Speed of Mode Evacuation Evacuation Evacuation Evacuation CAPB-1 4.66E-01 4.66E-01 4.67E-01 3.82E+06 3.82E+06 3.82E+06 CAPB-2 9.96E+01 9.97E+01 9.97E+01 6.40E+06 6.40E+06 6.40E+06 CAPB-3 1.06E+02 1.06E+02 1.06E+02 6.48E+06 6.48E+06 6.48E+06 CAPB-4 1.38E+04 1.39E+04 1.39E+04 4.28E+09 4.28E+09 4.28E+09 CAPB-5 1.81 E+04 1.82E+04 1.82E+04 5.30E+09 5.30E+09 5.30E+09 CAPB-6 1.51E+04 1.51E+04 1.51E+04 3.51E+09 3.51E+09 3.51E+09 CAPB-7 1.67E+04 1.68E+04 1.68E+04 4.42E+09 4.42E+09 4.42E+09 CAPB-8 4.1OE+04 4.16E+04 4.17E+04 1.47E+10 1.47E+10 1.47E+10 CAPB-9 2.37E+04 2.38E+04 2.39E+04 8.33E+09 8.33E+09 8.33E+09 CAPB-1D 4.31E+04 4.34E+04 4.36E+04 1.54E+1o 1.54E+10 1.54E+10 CAPB-11 3.45E+04 3.48E+04 3.49E+04 1.15E+10 1.15E+10 1.15E+10 CAPB-12 9.72E+01 9.75E+01 9.78E+01 4.63E+06 4.63E+06 4.63E+06 CAPB-13 7.30E+03 7.30E+03 7.31E+03 6.53E+08. 6.53E+08 6.53E+08 CAPB-14 1.58E+04 1.58E+04 1.59E+04 4.14E+09 4.14E+09 4.14E+09 CAPB-15 4.31E+04 4.33E+04 4.35E+04 1.59E+10 1.59E+10 1.59E+10 CAPB-16 1.86E+04 1.87E+04- 1.88E+04 5.50E+09 5.50E+09 5.50E+09 CAPB-17 4.81E+04 4.83E+04 4.86E+04 1.71E+10 1.71E+10 1.71E+10 CAPB-18 2.38E+04 2.39E4-04 2.40E+04 7.86E+09 7.86E+09 7.86E+09 CAPB-19 5.31E+04 5.33E+04 5.37E+04 1.88E+10 1.88E+10 1.88E+10 E.1-68

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.6 References E.1-1 ENN Engineering Report PNPS-PSA, "Pilgrim Nuclear Power Station Individual Plant Examination for Internal Events Update," April 2003, Revision 1.

E.1-2 Pilgrim Nuclear Power Station Individual Plant Examination, Revision 0, September 1992.

E.1-3 Boston Edison Company to the NRC, Response to Request for Additional Information Regarding the Pilgrim Individual Plant Examination (IPE) Submittal (TAC No. M74451, letter dated December 28, 1995 (2.95.127).

E.1-4 Modular Accident Analysis Program Boiling Water Reactor (MAAP BWR) Code, Version 4.0.4 and Fauske & Associates, Inc., "MAAP 4.0 Users manual," prepared for The Electric Power Research Institute, May 1994.

E.1-5 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.1-6 "Pilgrim Nuclear Power Station Individual Plant Examination for External Events," July 1994, Revision 0.

E.1-7 Parkinson, W. J., "EPRI Fire PRA Implementation Guide", prepared by Science Applications International Corporation for Electric Power Research Institute, EPRI TR-105928, December 1995.

E.1-8 U.S. Nuclear Regulatory Commission, NUREG-1 742, Perspectives Gained From the Individual Plant Examination of External Events (IPEEE) Program, Volume 1, Final Report, April 2002.

E.1-9 U.S. Nuclear Regulatory Commission, NUREG/CR-5500, Vol. 3, (INEEUEXT 00740), Reliability Study: General Electric Reactor Protection System, 1984-1995, May 1999.

E.1-10 Electric Power Research Institute, NSAC-154, "ISLOCA Evaluation Guidelines,"

prepared by ERIN Engineering and Research, Inc., September 1991.

E.1-11 Chu, et al., "Interfacing Systems LOCA: Boiling Water Reactors," Brookhaven National Laboratory, NUREG/CR-5124, BNL-NUREG-52141, February 1989.

E.1-12 Electric Power Research Institute, "PSA Applications Guide," EPRI TR-1 05396, prepared by ERIN Engineering and Research, Inc., August, 1995.

E.1-13 GOTHIC Containment Analysis Package, Version 3.4e, EPRI Tr-103053-V2, October 1993.

E.1-69

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1-14 U.S. Nuclear Regulatory Commission, NUREG/CR-5497, (INEEUEXT-97-01328),

Common-Cause Failure Parameter Estimations, October 1998.

E.1-15 Swain, A. D. and H. E. Guttmann, NUREG/CR-1 278, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications, Sandia National Laboratories, U.S. Nuclear Regulatory Commission, August 1983.

E.1-16 BWR Owners Group, "Pilgrim PSA Certification," BWROG/PSA-9903, March 2000.

E.1-17 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174 (draft was issued as DG-1061), "An Approach for Using Probabilistic Risk Assessment in Risk-informed Decisions on Plant-Specific Changes to the Licensing Basis," July 1998.

E.1-18 Chanin, D. I., and M. L. Young, Code Manual for MACCS2: Volume 1, User's Guide, SAND97-0594 Sandia National Laboratories, Albuquerque, NM, 1997.

E.1-19 Boston Edison Company, "Appendix I Evaluation," forwarding evaluation of Pilgrim Station Unit 1 Conformance to the Design Objectives of 10 CFR 50, Appendix I, letter dated March 31, 1977 (2.77.031).

E.1-20 PNPS Emergency Plan, Revision 24, February 7, 2001, Appendix 5, Pilgrim Station Evacuation Time Estimates and Traffic Management Plan Update, Revision 5, November 1998.

E.1-21 U.S. Nuclear Regulatory Commission, NUREG/CR-4551, Vol. 2, Rev. 1, Part 7, Evaluation of Severe Accident Risks: Quantification of Major Input Parameters, MACCS Input, December 1990.

E.1-70

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage ATTACHMENT E.2 SAMA CANDIDATES SCREENING AND EVALUATION QW

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.2 EVALUATION OF 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 A list of SAMA candidates was developed by reviewing industry documents and considering plant-specific enhancements not identified in published industry documents. Since PNPS is a conventional GE nuclear power reactor design, considerable attention was paid to the SAMA candidates from SAMA analyses for other GE plants. Industry documents reviewed include the following:

  • Hatch SAMA Analysis (Reference E.2-1),
  • Calvert Cliffs Nuclear Power Plant SAMA Analysis (Reference E.2-2),
  • GE ABWR SAMDA Analysis (Reference E.2-3),
  • Peach Bottom SAMA Analysis (Reference E.2-4),
  • Quad Cities SAMA Analysis (Reference E.2-5),
  • Dresden SAMA Analysis (Reference E.2-6), and

The above documents represent a compilation of most SAMA candidates developed from the industry documents. These sources of other industry documents include the following:

  • Limerick SAMDA cost estimate report (Reference E.2-8),
  • NUREG-1437 description of Limerick SAMDA (Reference E.2-9),
  • NUREG-1437 description of Comanche Peak SAMDA (Reference E.2-1 0),
  • Watts Bar SAMDA submittal (Reference E.2-11),
  • TVA's response to NRC's RAI on the Watts Bar SAMDA submittal (Reference E.2-12),
  • Westinghouse AP600 SAMDA (Reference E.2-13),
  • NUREG-0498, Watts Bar Final Environmental Statement Supplement 1, Section 7 (Reference E.2-14),
  • NUREG-1 560, Volume 2, NRC Perspectives on the IPE Program (Reference E.2-15),

and

  • NUREG/CR-5474, Assessment of Candidate Accident Management Strategies (Reference E.2-16).

E.2-1

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage In addition to SAMA candidates from review of industry documents, additional SAMA candidates were obtained from plant-specific sources, such as the PNPS IPE (Reference E.2-17) and IPEEE (Reference E.2-18). In both 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 numbers 248 through 281. The current PNPS PSA model was also used to identify plant-specific modifications for inclusion in the comprehensive list of SAMA candidates. The risk-significant terms 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 were listed in Table E.1-2.

The comprehensive list, available in on-site documentation, contained a total of 281 phase I SAMA candidates.

E.2.2 Qualitative Screenina 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 PNPS. Potential SAMA candidates were screened out if they modified features not applicable to PNPS, if they had already been implemented at PNPS, 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, 63 of the phase I SAMA candidates were screened out because they were not applicable to PNPS, 4 of the phase I SAMA candidates were screened out because they were similar in nature and could be combined with another SAMA candidate, and 155 of the phase I SAMA candidates were screened out because they had already been implemented at PNPS, leaving 59 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.2-1 provides a description of each of the 59 phase 11SAMA 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 SAMA and SAMDA analyses. In particular, these cost-estimates were derived from the following major sources:

  • GE ABWR SAMDA Analysis (Reference E.2-3),
  • Peach Bottom SAMA Analysis (Reference E.2-4),

E.2-2

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage

  • Quad Cities SAMA Analysis (Reference E.2-5),
  • Dresden SAMA Analysis (Reference E.2-6),
  • ANO-2 SAMA Analysis (Reference E.2-7), and The cost estimates did not include the cost of replacement power during extended outages required to implement the modifications, nor did they include contingency costs associated with unforeseen implementation obstacles. Estimates based on modifications that were implemented or estimated in the past were presented in terms of dollar values at the time of implementation (or estimation), and were not adjusted to present-day dollars. In addition, several implementation costs were originally developed for SAMDA analyses (i.e., during the design phase of the plant),

and therefore, do not capture the additional costs associated with performing design modifications to existing plants (i.e., reduced efficiency, minimizing dose, disposal of contaminated material, etc.). Therefore, the cost estimates were conservative.

The benefit of implementing a SAMA candidate was estimated in terms of averted consequences. 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 NUREG/BR-0184 (Reference E.2-19) 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.

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. Several of the SAMA candidates were clearly in excess of the attainable benefit estimated from a particular analysis case.

For less clear cases, engineering judgment on the cost associated with procedural changes, engineering analysis, testing, training, and hardware modification was applied to determine if a more detailed cost estimate was necessary to formulate a conclusion regarding the economic viability of a particular SAMA. Based on a review of previous submittals' SAMA evaluations and an evaluation of expected implementation costs at PNPS, the following estimated costs for each potential element of the proposed SAMA implementation are used.

E.2-3

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Type of Change Estimated Cost Range Procedural only $25K-$50K Procedural change with engineering $50K-$200K required Procedural change with engineering and $200K-$300K testing/training required Hardware modification $1OOK to >$1OOOK In most cases, more detailed cost estimates were not required, particularly if the SAMA called for the implementation of a hardware modification. Nonetheless, the cost of each unscreened SAMA candidate was conceptually estimated to the point where conclusions regarding the economic viability of the proposed modification could be adequately gauged. The cost benefit comparison and disposition of each of the 59 phase 11SAMA candidates is presented in Table E.2-1.

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 a digital large break LOCA protection system. The bounding calculation estimated the benefit of this improvement by total elimination of risk due to large break LOCA (see analysis in phase 11SAMA 052 of Table E.2-1). 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.

Decay Heat Removal Capability - Torus Cooling This analysis case was used to evaluate the change in plant risk from installing an additional decay heat removal 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 the torus cooling mode of the RHR system to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $261,832. This analysis case was used to model the benefit of phase 11SAMAs 1 and 14.

Decay Heat Removal Capability - Drywell Sp=ra This analysis case was used to evaluate the change in plant risk from installing an additional decay heat removal system. Enhancements of decay heat removal capability decrease the E.2-4

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage probability of loss of containment heat removal. A bounding analysis was performed by setting the events for loss of the drywell spray mode of the RHR system to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $264,219. This analysis case was used to model the benefit of phase 11SAMA 9.

Filtered Vent This analysis case was used to evaluate the change in plant risk from installing a filtered containment vent to provide fission product scrubbing. A bounding analysis was performed by reducing the successful torus venting accident progression source terms by a factor of 2 to reflect the additional filtered capability. Reducing the releases from the vent path resulted in no benefit. This analysis case was used to model the benefit of phase 11SAMAs 2 and 19.

Containment Vent for ATWS Decay Heat Removal This analysis case was used to evaluate the change in plant risk from installing a containment vent to provide alternate decay heat removal capability during an ATWS event. A bounding analysis was performed by setting the ATWS sequences associated with containment bypass to zero in the level I PSA model, which resulted in an upper bound benefit of approximately

$61,701. This analysis case was used to model the benefit of phase 11SAMAs 3 and 47.

Molten Core Debris Removal This analysis case was used to estimate the change in plant risk from providing a molten core debris cooling mechanism. A bounding analysis was performed by setting containment failure due to core-concrete interaction (not including liner failure) to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately $2,620,551. This analysis case was used to model the benefit of phase 11SAMAs 4, 5, 8, and 23.

Dryweff Head Flooding This analysis case was used to evaluate the change in plant risk from providing a modification to flood the drywell head such that if high drywell temperature occurred, the drywell head seal would not fail. A bounding analysis was performed by setting the probability of drywell head failure due to high temperature to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately $12,915. This analysis case was used to model the benefit of phase 11SAMAs 6,18, and 20.

Reactor Building Effectiveness This analysis case was used to evaluate the change in plant risk by ensuring the reactor building is available to provide effective fission product removal. Reactor building effectiveness was conservatively modeled by assuming reactor building availability for all accident sequences. This resulted in an upper bound benefit of approximately $64,577. This analysis case was used to model the benefit of phase II SAMAs 7, 13, and 21.

E.2-5

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Strengthen Containment This analysis case was used to evaluate the change in plant risk from strengthening containment to reduce the probability of containment over-pressurization failure. A bounding analysis was performed by setting all energetic containment failure modes (DCH, steam explosions, late over-pressurization) to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately $1,233,428. This analysis case was used to model the benefit of phase 11SAMAs 10, 15, 16, and 24.

Base Mat Melt-Through This analysis case was used to evaluate the change in plant risk from increasing the depth of the concrete base mat to ensure base mat melt-through does not occur. A bounding analysis was performed by setting containment failure due to base mat melt-through to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately $25,831. This analysis case was used to model the benefit of phase 11SAMA 11.

Reactor Vessel Exterior Coolinm This analysis case was used to evaluate the change in plant risk from providing a method to perform ex-vessel cooling of the lower reactor vessel head. A bounding analysis was performed by modifying the probability of vessel failure by a factor of two to account for ex-vessel cooling in the level 2 PSA model, which resulted in an upper bound benefit of approximately $19,373. This analysis case was used to model the benefit of phase 11SAMA 12.

Vacuum Breakers This analysis case was used to evaluate the change in plant risk from improving the reliability of vacuum breakers to reseat following a successful opening and eliminate suppression pool scrubbing failures from the containment analysis. A bounding analysis was performed by setting the vacuum breaker failure probability to zero in the level 1 PSA model, which resulted in no benefit. This analysis case was used to model the benefit of phase 11SAMA 17.

Flooding the Rubble Bed This analysis case was used to evaluate the change in plant risk from providing a source of water to the drywell floor to flood core debris. A bounding analysis was performed by substituting the probabilities of wet core concrete interactions for dry core concrete interactions in the level 2 PSA model, which resulted in an upper bound benefit of approximately $1,226,971. This analysis case was used to model the benefit of phase 11SAMA 22.

DC Power This analysis case was used to evaluate the change in plant risk from plant modifications that would increase the availability of Class 1E DC power (e.g., increasing battery capacity, using fuel cells, or extending SBO injection provisions). It was assumed that battery life could be extended E.2-6

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage from 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to simulate additional battery capacity. This enhancement would extend HPCI and RCIC operability and allow more credit for AC power recovery. A bounding analysis was performed by changing the time available to recover offsite power before HPCI and RCIC are lost from 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during SBO scenarios in the level I PSA model. This resulted in an upper bound benefit of approximately $146,356. This analysis case was used to model the benefit of phase 11SAMAs 25, 26, 28, 33, and 35.

Improve DC System This analysis case was used to evaluate the change in plant risk from improving injection capability by auto-transfer of AC bus control power to a standby DC power source upon loss of the normal DC source or from enhancing procedure to make use of DC bus cross-tie to improve DC power availability and reliability. A bounding analysis was performed by setting the DC buses D1 6 and D1 7 to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $118,568. This analysis case was used to model the benefit of phase 11SAMAs 27 and 34.

Altemate Pump Power Source This analysis case was used to evaluate the change in plant risk from adding a small, dedicated power source such as a dedicated diesel or gas turbine for the feedwater or condensate pumps so that they do not rely on offsite power. A bounding analysis was performed by setting failure of the SBO diesel generator to zero in level 1 PSA model, which resulted in an upper bound benefit of approximately $265,687. This analysis case was used to model the benefit of phase 11SAMA 29.

Improve AC Power System This analysis case was used to evaluate the change in plant risk from improving AC power system cross-tie capability to enhance the availability and reliability of the AC power system. A bounding analysis was performed by setting the loss of MCCs B17, B18, and B15 to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $473,410. This analysis case was used to model the benefit of phase 11SAMA 30.

Dedicated DC Power and Additional Batteries and Divisions This analysis case was used to evaluate the change in plant risk from plant modifications that would provide motive power to components (e.g., providing a dedicated DC power supply, additional batteries, or additional divisions). A bounding analysis was performed by setting the loss of DC bus D17 initiator, and one division of DC power, to zero in the level 1 PSA model, which resulted In an upper bound benefit of approximately $903,025. This analysis case was used to model the benefit of phase 11SAMAs 31 and 32.

E.2-7

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Locate RHR Inside Containment This analysis case was used to evaluate the change in plant risk from moving the RHR system inside containment to prevent an RHR system ISLOCA event outside containment. A bounding analysis was performed by setting the RHR ISLOCA sequences to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $16,497. This analysis case was used to model the benefit of phase 11SAMA 36.

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. A bounding analysis was performed by setting the ISLOCA initiator to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $24,148. This analysis case was used to model the benefit of phase 11SAMA 37.

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 setting the containment bypass failure due to MSIV leakage to zero in the level 2 PSA model, which resulted in no benefit. This analysis case was used to model the benefit of phase 11SAMA 38.

Diesel to CST Makeup Pumps This analysis case was used to evaluate the change in plant risk from installing an independent diesel for the CST makeup pumps to allow continued operation of the high pressure injection system during an SBO event. As currently modeled, if CST water level is low, swapping HPCI/

RCIC suction from the CST to the torus allows continued HPCI and RCIC injection. Therefore, a bounding analysis was performed by setting the failure to switchover from CST to torus to zero in the level 1 PSA model, which resulted in no benefit. This analysis case was used to model the benefit of phase 11SAMA 39.

High Pressure Injection System This analysis case was used to evaluate the change in plant risk from plant modifications that would increase the availability of high pressure injection (e.g., installing an independent AC powered high pressure injection system, passive high pressure injection system, or an additional high pressure injection system). A bounding analysis was performed by setting the CDF contribution due to unavailability of the HPCI system to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $110,212. This analysis case was used to model the benefit of phase II SAMAs 40, 41, 42, 44, and 45.

E.2-8

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Improve the Reliability of High Pressure Injection System This analysis case was used to evaluate the change in plant risk from plant modifications that would increase the reliability of the high pressure injection system. A bounding analysis was performed by reducing the HPCI system failure probability by a factor of three in the level 1 PSA model, which resulted in an upper bound benefit of approximately $76,025. This analysis case was used to model the benefit of phase 11SAMA 43.

SRVs Reseat This analysis case was used to evaluate the change in plant risk from improving the reliability of SRVs reseating. A bounding analysis was performed by setting the stuck open SRVs initiator to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately

$63,599. This analysis case was used to model the benefit of phase 11SAMA 46.

Diversity of Explosive Valves This analysis case was used to evaluate the change in plant risk from providing an alternate means of opening a pathway to the RPV for SLC system injection, thereby improving success probability for reactor shutdown. A bounding analysis was performed by setting common cause failure of SLC explosive valves to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $12,915. This analysis case was used to model the benefit of phase II SAMA48.

Reliability of SRVs This analysis case was used to evaluate the change in plant risk from installing additional signals to automatically open the SRVs. This improvement would reduce the likelihood of SRVs failing to open, thereby reducing the consequences of medium LOCAs. A bounding analysis was performed by setting the probability of SRVs failing to open when required by reactor pressure vessel overpressure conditions to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $31,799. This analysis case was used to model the benefit of phase 11SAMA 49.

Improve SRV Design This analysis case was used to evaluate the change in plant risk from improving the SRV design to increase the reliability of opening, thus increasing the likelihood that accident sequences could be mitigated using low pressure injection systems. A bounding analysis was performed by setting the probability of SRVs failing to open during RPV depressurization to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $194,378. This analysis case was used to model the benefit of phase 11SAMA 50.

E.2-9

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Self-Cooled ECCS Pump Seals This analysis case was used to evaluate the change in plant risk from providing self-cooled ECCS pump seals to eliminate dependence on the component cooling water system. A bounding analysis was performed by setting the CDF contribution from sequences involving RHR pump failures to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $29,412. This analysis case was used to model the benefit of phase 11SAMA 51.

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 break LOCA initiator to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately $14,109. This analysis case was used to model the benefit of phase 11SAMA 52.

Controlled Containment Venting This analysis case was used to evaluate the change in plant risk from changing the design of the containment vent valves and procedure to establish a narrow pressure control band. This would prevent rapid containment depressurization when venting, thus avoiding adverse impact on the ability of the low pressure ECCS injection systems to take suction from the torus. A bounding analysis was performed by reducing the probability of the operator failing to recognize the need to vent the torus by a factor of three in the level 1 PSA model, which resulted in an upper bound benefit of approximately $137,237. This analysis case was used to model the benefit of phase 11 SAMA 53.

ECCS Low Pressure Interlock This analysis case was used to evaluate the change in plant risk from installing a bypass switch to allow operator to bypass the ECCS low pressure interlock circuitry that inhibits opening of the RHR low pressure injection and core spray injection valves following sensor or logic failure. A bounding analysis was performed by setting the CDF contribution due to sensor failure, low pressure permissive logic failure, and miscalibration to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately $21,761. This analysis case was used to model the benefit of phase 11SAMA 54.

Improve the Reliability of SSW and RBCCW Pumps This analysis case was used to evaluate the change in plant risk from providing a separate pump train to eliminate common cause failure of SSW and RBCCW pumps. A bounding analysis was performed by setting the CDF contribution due to common cause failures of SSW and RBCCW pumps to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately $356,310. This analysis case was used to model the benefit of phase 11SAMA 55.

E.2-10

  • Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Redundant DC Power Supplies to DTV Valves This analysis case was used to evaluate the change in plant risk from installing additional fuses to two DTV valve control circuits to enable the DTV function. A bounding analysis was performed by setting the CDF contribution due to DC power supply failures to DTV valves AO-5042B and AO-5025 to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately $220,639. -This analysis case was used to model the benefit of phase 11SAMA 56.

Proceduralize the Use of Diesel Fire Pump Hydroturbine This analysis case was used to evaluate the change in plant risk from revising the procedure to allow use of hydroturbine if EDG X-107A or diesel driven fire water pump P-140 is unavailable. A bounding analysis was performed by setting the CDF contribution from the sequences involving a LOOP and failure of either EDG A or fuel oil transfer oil pump (P-141) to zero in the level I PSA model. This resulted in an upper bound benefit of approximately $175,279. This analysis case was used to model the benefit of phase 11SAMA 57.

Proceduralize Alignment of Bus B3 to Feed Bus BI Loads or Bus B4 to Bus B2 This analysis case was used to evaluate the change in plant risk from providing a procedure to direct the operator to restore 480V MCCs B15 and B17 loads upon loss of 4.16kV bus A5 provided that 4.16kV bus A3 is available. The same is true for restoring 480V MCCs B14 and B18 loads upon loss of 4.16kV bus A6 provided that 4.16kV bus A4 is available. A bounding analysis was performed by setting the CDF contribution from the sequences involving a loss of the 4.16 kV bus A5 to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately $190,797. This analysis case was used to model the benefit of phase 11SAMA 58.

Redundant Path from Fire Water Pump Discharge to LPCI Loops A and B Cross-tie This analysis case was used to evaluate the change in plant risk from installing a redundant path from fire protection water pump discharge to LPCI loops A and B cross-tie. A bounding analysis was performed by setting the CDF contribution from the sequences involving fire water into LPCI loops A and B cross-tie failure to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately $929,797. This analysis case was used to model the benefit of phase 11 SAMA 59.

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.2-2.

A description of each sensitivity case follows.

E.2-11

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's 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 27-year period for remaining plant life (i.e. seven years on the original plant license plus the 20-year license renewal period). The 20-year license renewal period was used in the base case. The resultant monetary equivalent was calculated using 27 years remaining until end of facility life to investigate the impact on each analysis case. Changing this assumption does not cause any 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 any additional SAMAs to be cost-beneficial.

E.2-12

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.2.5 References E.2-1 Appendix D-Attachment F,Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Edwin I. Hatch Nuclear Power Plant Units 1 and 2, March 2000.

E.2-2 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Calvert Cliffs Nuclear Power Plant, Supplement 1, February 1999.

E.2-3 General Electric Nuclear Energy, Technical Support Document for the ABWR, 25A5680, Revision 1, January 18,1995.

E.24 Appendix E- Environmental Report, Appendix G, Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Peach Bottom Nuclear Power Plant Units 2 and 3, July, 2001.

E.2-5 Appendix F,Severe Accident Mitigation Alternatives Analysis Submittal Related to Licensing Renewal for the Quad Cities Nuclear Power Plant Units 1 and 2, January 2003.

E.2-6 Appendix F,Severe Accident Mitigation Alternatives Analysis Submittal Related to Licensing Renewal for the Dresden Nuclear Power Plant Units 2 and 3, January 2003.

E.2-7 Appendix E-Attachment E, Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Arkansas Nuclear One - Unit 2, October 2003.

E.2-8 Cost Estimate for Severe Accident Mitigation Design Alternatives, Limerick Generating Station for Philadelphia Electric Company, Bechtel Power Corporation, June 22, 1989.

E.2-9 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Volume 1, 5.35, Listing of SAMDAs considered for the Limerick Generating Station, May 1996.

E.2-10 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Volume 1, 5.36, Listing of SAMDAs considered for the Comanche Peak Steam Electric Station, May 1996.

E.2-11 Museler, W. J., (Tennessee Valley Authority) to NRC Document Control Desk, "Watts Bar Nuclear Plant (WBN) Units I and 2 - Severe Accident Mitigation Design Alternatives (SAMDAs)," letter dated October 7, 1994.

E.2-12 Nunn, D. E., (TVA) to NRC Document Control Desk, "Watts Bar Nuclear Plant (WBN)

Units I and 2 - Severe Accident Mitigation Design Alternatives (SAMDA) - Response to Request for Additional Information (RAI) - (TAC Nos. M77222 and M77223)," letter dated October 7, 1994.

E.2-13

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.2-13 Liparulo, N. J., (Westinghouse Electric Corporation) to NRC Document Control Desk, "Submittal of Material Pertinent to the AP600 Design Certification Review," letter dated December 15,1992.

E.2-14 U.S. Nuclear Regulatory Commission, NUREG-0498, Final Environmental Statement related to the operation of Watts Bar Nuclear Plant, Units 1 and 2, Supplement No. 1, April 1995.

E.2-15 U.S. Nuclear Regulatory Commission, NUREG-1 560, Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance, Volume 2, December 1997.

E.2-16 U.S. Nuclear Regulatory Commission, NUREG/CR-5474, Assessment of Candidate Accident Management Strategies, March 1990.

E.2-17 Pilgrim Nuclear Power Station, Individual Plant Examination (IPE) Report, September 1992 E.2-18 Pilgrim Nuclear Power Station, Individual Plant Examination of External Events (IPEEE)

Report, July 1994.

E.2-19 U.S. Nuclear Regulatory Commission, NUREG/BR-01 84, Regulatory Analysis Technical Evaluation Handbook, January 1997.

QW E.2-14

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding t SAMA Analysis .00 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Phase II l Result of Potential SAMA ID SAMA Enhancement Improvements Related to Accident Mitigation Containment Phenomena 001 Install an SAMA would decrease 4.70% 4.60% $43,639 $261,832 $5,800,000 Not cost independent the probability of loss effective method of of containment heat suppression pool removal.

cooling. I Basis for

Conclusion:

The CDF contribution from loss of the torus cooling mode of RHR was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $5.8 million. Therefore, this SAMA is not cost effective for PNPS.

002 Install a filtered SAMA would provide 0.00% 0.00% $0 $0 $3,000,000 Not cost containment vent an alternate decay effective to provide fission heat removal method product for non-ATWS events, scrubbing. with fission product Option 1: Gravel scrubbing.

Bed Filter Option 2: Multiple Venturi Scrubber Basis for

Conclusion:

Successful torus venting accident progression source terms are reduced by a factor of 2 to reflect the additional filtered capability. The cost of implementing this SAMA at Peach Bottom was estimated to be $3 million. Therefore, this SAMA is not cost effective for PNPS.

E.2-15

9.~ 3 NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase II Result of Potential CDF off-Site Estimated Upper Estimated BoundEsti ated Conclusion SAMA ID Enhancement Reduction dos Benefit Estimated Cost ReductionBnet 003 Install a Assuming that injection 0.50% 1.19% $10,283 $61,701 >$2,000,000 Not cost containment vent isavailable, this SAMA effective large enough to would provide alternate remove ATWS decay heat removal in decay heat. an ATWS event.

Basis for

Conclusion:

The CDF contribution from ATWS sequences associated with containment bypass were eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million.

Therefore, this SAMA is not cost effective for PNPS.

004 Create a large SAMA would ensure 0.00% 48.62% $436,759 $2,620,551 >$100 million Not cost concrete crucible that molten core debris effective with heat removal escaping from the potential under vessel would be the base mat to contained within the contain molten crucible. The water core debris. cooling mechanism would cool the molten core, preventing a melt-through of the base mat.

Basis for

Conclusion:

Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at ANO-2 was estimated to be $100 million.

Therefore, this SAMA is not cost effective for PNPS.

E.2-16

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase II SAMA Result of Potential SAMA ID Enhancement

.i 005 Create a water- SAMA would contain cooled rubble bed molten core debris on the pedestal. dropping on to the pedestal and would allow the debris to be cooled.

Basis for

Conclusion:

Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at ANO-2 was estimated to be $19 million.

Therefore, this SAMA is not cost effective for PNPS.

006 Provide SAMA would provide 0.00% 0.07% $2,153 $12,915 >$1,000,000 Not cost modification for intentional flooding of effective flooding the the upper drywell head drywell head. such that if high drywell temperatures occurred, the drywell head seal would not fail.

Basis for

Conclusion:

Drywell head failures due to high temperature were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $1 million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

E.2-17

Exhibit No. NRC000001 J 3 NRC - Applicant's Environmental Report SAMA Analysis J Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase II Result of Potential CDF Off-Site Estimated Upper Estimated SAMA ID Enhancement Reduction Reduction Benefit Estimated Cost RedutionBenefit 007 Enhance fire SAMA would improve 0.00% 1.16% $10,763 $64,577 >$2,500,000 Not cost protection system fission product effective and SGTS scrubbing in severe hardware and accidents.

procedures.

Basis for

Conclusion:

Failure of the reactor building to contain releases was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2.5 million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

E.2-18

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR t_' V ir_'

Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued) off-site Upper Phase II Result of Potential CDF Estimated Bound Estimated SAMA ID Enhancement Reduction Benefit Estimated Cost Conclusion

~: ReductionBefi Benefit 008 Create a core melt SAMA would provide 0.00% 48.62% $436,759 $2,620,551 >$5,000,000 Not cost source reduction cooling and effective system. containment of molten core debris. Refractory material would be placed underneath the reactor vessel such that a molten core falling on the material would melt and combine with the material. Subsequent spreading and heat removal from the vitrified compound would be facilitated, and concrete attack would not occur.

Basis for

Conclusion:

Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $5 million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

E.2-19

3 I3__

NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 J,

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase II Result of Potential CDF OffEstimate Estmaedd Upper Estimated Bondostmaed Conclusion SAMA ID SAMA Enhancement Reduction Reduce on Benefit Estimated Cost RedutionBenefit 009 Install a passive SAMA would decrease 5.05% 4.70% $44,037 $264,219 $5,800,000 Not cost containmentspray the probability of loss effective system. of containment heat removal.

Basis for

Conclusion:

The CDF contribution from loss of the drywell spray mode of RHR was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $5.8 million. Therefore, this SAMA is not cost effective for PNPS.

010 Strengthen SAMA would reduce 0.00% 26.10% $205,571 $1,233,428 $12,000,000 Not cost primary and the probability of effective secondary containment over-containment. pressurization failure.

Basis for

Conclusion:

Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR was estimated to be $12 million. Therefore, this SAMA is not cost effective for PNPS.

E.2-20

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase II Result of Potential CDF lt Estimated Estmatd und BunoEsimaedEstimated Conclusion SAMA ID Enhancement Reduction Reduction Benefit Estimated Cost RedutionBenefit 011 Increase the SAMA would prevent 0.00% 0.43% $4,305 $25,831 >$5,000,000 Not cost depth of the base mat melt-through. effective concrete base mat or use an alternative concrete material to ensure melt-through does not occur. l Basis for

Conclusion:

Containment failure due to base mat melt-through was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $5 million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

012 Provide a reactor SAMA would provide 0.00% 0.22% $3,229 $19,373 $2,500jO00 Not cost vessel exterior the potential to cool a effective cooling system. molten core before it causes vessel failure, if the lower head could be submerged in water.

Basis for

Conclusion:

The probability of vessel failure was modified to account for potential ex-vessel cooling of the vessel bottom head region to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $2.5 million. Therefore, this SAMA is not cost effective for PNPS.

E.2-21

J 3j NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)

Phase 11 Result of Potential CDF Offite Estimated Est ma ed Bound Bo ndost Estimated ma ed Conclusion SAMA ID M Enhancement Reduction Reduction Benefit Estimated Cost Redu tionBenefit 013 Construct a SAMA would provide a 0.00% 1.16% $10,763 $64,577 >$2,000,000 Not cost building method to effective connected to depressurize primary containment and containment that reduce fission product is maintained at a release.

vacuum.

Basis for

Conclusion:

Failure of the reactor building to contain releases was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.

014 2.g. Dedicated SAMA would decrease 4.70% 4.60% $43,639 $261,832 $5,800,000 Not cost Suppression Pool the probability of loss effective Cooling of containment heat removal.

Basis for

Conclusion:

The CDF contribution from loss of the torus cooling mode of RHR was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $5.8 million. Therefore, this SAMA is not cost effective for PNPS.

015 3.a. Create a SAMA increases time 0.00% 26.10% $205,571 $1,233,428 $8,000,000 Not cost larger volume in before containment effective containment. failure and increases time for recovery.

Basis for

Conclusion:

Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $8 million.

Therefore, this SAMA is not cost effective for PNPS.

E.2-22

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding I : -A I:0 SAMA Analysis 50-293-LR, 06-848-02-LR IK-:

Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase II Result of Potential CDF Off-Site Upper Bound EstimaEsE SAMA Esimtdoond Esiatd Conclusion SAMA ID Enhancement Reduction Benefit Estimated Cost:

ReductionBefi 016 3.b. Increase SAMA minimizes 0.00% 26.10% $205,571 $1,233,428 $12,000,000 Not cost containment likelihood of large effective pressure releases.

capability (sufficient pressure to withstand severe accidents).

Basis for

Conclusion:

Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR was estimated to be $12 million. Therefore, this SAMA is not cost effective for PNPS.

017 3.c. Install This SAMA addresses 0.00% 0.00% $0 $0 >$1,000,000 Not cost improved vacuum the reliability of a effective breakers vacuum breaker to (redundant valves reseat following a in each line). successful opening.

Basis for

Conclusion:

Vacuum breaker failures and suppression pool scrubbing failures were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $1 million.

Therefore, this SAMA is not cost effective for PNPS.

E.2-23

J) 3 NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Off- itesti m ted Upper Phase 11 Result of Potential CDF OffSite Estimated Und Estimated SAMA ID Enhancement Reduction Rducion Benefit Esim d Cost Redu tionBenefit 018 3.d. Increase the This SAMA would 0.00% 0.07% $2,153 $12,915 $12,000,000 Not cost temperature reduce the potential for effective margin for seals. containment failure under adverse conditions.

Basis for

Conclusion:

Containment failure due to high temperature drywell seal failure was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR were estimated to be $12 million and was judged to exceed the attainable benefit, even without a detailed cost estimate. Therefore, this SAMA is not cost effective for PNPS.

019 5.b/c. Install a SAMA would provide 0.00% 0.00% $0 $0 $3,000,000 Not cost filtered vent an alternate decay effective heat removal method for non-ATWS events, with fission product scrubbing.

Basis for

Conclusion:

Successful torus venting accident progressions source terms are reduced by a factor of 2 to reflect the additional filtered capability. The cost of implementing this SAMA at Peach Bottom was estimated to be $3 million. Therefore, this SAMA is not cost effective for PNPS.

E.2-24

Exhibit No. NRC000001 NRC - Applicant's I-' Environmental Report Pilgrim LR Proceeding 11 delo, n SAMA Analysis 50-293-LR, 06-848-02-LR 1_,

Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase II Result of Potential CDF i Estimate Estmaedd BuEstimated Bondostmaed

Conclusion:

SAMA ID Enhancement Reduction De Benefit Estimated Cost Reduction Benefit 020 7.a. Provide a SAMA would provide 0.00% 0.07% $2,153 $12,915 >$1,000,000 Not cost method of drywell intentional flooding of effective head flooding. the upper drywell head such that if high drywell temperatures occurred, the drywell head seal would not fail.

Basis for

Conclusion:

Drywell head failures due to high temperature were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $1 million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

021 13.a. Use This SAMA provides 0.00% 1.16% $10,763 $64,577 >$2,500,000 Not cost alternate method the capability to use effective of reactor building firewater sprays in the spray. reactor building to mitigate release of fission products into the reactor building following an accident.

Basis for

Conclusion:

Failure of the reactor building to contain releases was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2.5 million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

E.2-25

3 J NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR I

Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase II Result of Potential CDF OffSte Estimated Est ma ed Bound Bo ndost Estimated ma ed Conclusion SAMA ID SAMA Enhancement Reduction Rductios Benefit Estimated Cost Redu tionBenefit 022 14.a. Provide a SAMA would allow the 0.00% 22.48% $204,495 $1,226,971 $2,500,000 - Not cost means of flooding debris to be cooled. effective the rubble bed.

Basis for

Conclusion:

The probabilities of wet core concrete interactions were substituted for dry core concrete interactions to assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $2.5 million. Therefore, this SAMA is not cost effective for PNPS.

023 14.b. Install a SAMA would enhance 0.00% 48.62% $436,759 $2,620,551 $8,750,000 Not cost reactor cavity debris coolability, effective flooding system. reduce core concrete interaction, and provide fission product scrubbing.

Basis for

Conclusion:

Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at ANO-2 was estimated to be $8.75 million.

Therefore, this SAMA is not cost effective for PNPS.

024 Add ribbing to the This SAMA would 0.00% 26.10% $205,571 $1,233,428 $12,000,000 Not cost containment shell. reduce the chance of effective containment buckling under reverse pressure loading.

Basis for

Conclusion:

Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR was estimated to be $12 million. Therefore, this SAMA is not cost effective for PNPS.

E.2-26

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase II Result of Potential CDF Off-Site Estimated Estmaed Upper BondostmaedEstimated Conclusion SAMA ID SAMA Enhancement Reduction Benefit Estimated Cost Improvements Related to Enhanced AC/DC Reliability/Availability R 025 Provide additional SAMA would ensure 1.39% 2.79% $24,393 $146,356 $500,000 Not cost DC battery longer battery effective capacity. capability during an SBO, which would extend HPCW/RCIC operability and allow more time for AC power recovery.

Basis for

Conclusion:

The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

026 Use fuel cells SAMA would extend 1.39% 2.79% $24,393 $146,356 >$2,000,000 Not cost instead of lead- DC power availability in effective acid batteries. an SBO, which would extend HPCI/RCIC operability and allow more time for AC power recovery.

Basis for

Conclusion:

The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2million. Therefore, this SAMA is not cost effective for PNPS.

E.2-27

Exhibit No. NRC000001 J SAMA Analysis 31 NRC - Applicant's Environmental Report Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase 11 Result of Potential CDF OffSite Estimated Bound Estimated SAMA Esim tdoo ndEsi Costatd Conclusion SAMA ID Enhancement Reduction Rduction Benefit Estimated Redu tionBenefit 027 Modification for SAMA would increase 4.65% 1.91% $19,761 $118,568 $500,000 Not cost Improving DC Bus reliability of AC power effective Reliability and injection capability.

Basis for

Conclusion:

The CDF contribution due to loss of DC buses D16 and D07 was eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment. Therefore, this SAMAis not cost effective for PNPS.

028 2.i. Provide 16- SAMA includes 1.39% 2.79% $24,393 $146,356 $500,000 Not cost hour SBO improved capability to effective injection. cope with longer SBO l scenarios. l Basis for

Conclusion:

The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

E.2-28

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)

Phase II Result of Potential CDF OffEstimate Estmaedd Bound Estimated Bondostmaed Conclusion SAMA ID SAM Enhancement Reduction Benefit Estimated Cost ReductionBefi 029 9.b. Provide an This SAMA would 2.22% 5.06% $44,281 $265,687 >$2,000,000 Not cost alternate pump provide a small, effective power source. dedicated power source such as a dedicated diesel or gas turbine for the feedwater or condensate pumps so that they do not rely on offsite power.

Basis for

Conclusion:

The CDF contribution due to failure of the SBO diesel was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.

030 9.g. Enhance SAMA would provide 11.10% 8.47% $78,902 $473,410 $146,120 Retain procedures to increased reliability of make use of AC AC power system and bus cross-ties. reduce core damage and release frequencies.

Basis for

Conclusion:

The CDF contribution due to loss of MCCs B17, B18, and B15 was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $146,120 by engineering judgment.

E.2-29

I-) 3 NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001

.

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase 11 SAMA Result of Potential CDF Off-SiteUpe ose Estimated Bound Estimated SAMA ID Enhancement Reduction ReductionBeet Benefit Estimated Cost 031 10.a. Add a This SAMA addresses 24.3% 16.16% $150,504 $903,025 $3,000,000 Not cost dedicated DC the use of a diverse DC effective power supply. power system such as an additional battery or fuel cell for the purpose of providing motive power to certain components (e.g.,

RCIC).

Basis for

Conclusion:

The CDF contribution due to loss of DC Bus 'B' was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $3million. Therefore, this SAMA is not cost effective for PNPS.

032 10.b. Install This SAMA addresses 24.3% 16.16% $150,504 $903,025 $3,000,000 Not cost additional the use of a diverse DC effective batteries or power system such as divisions. an additional battery or fuel cell for the purpose of providing motive power to certain components (e.g.,

RCIC).

Basis for

Conclusion:

The CDF contribution due to loss of DC Bus 'B' was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $3 million. Therefore, this SAMA is not cost effective for PNPS.

E.2-30

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding Ar-SAMA Analysis 50-293-LR, 06-848-02-LR VI I-Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase I SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)

Phase II Result of Potential CDF Off-Site Estimated Bound Estimated SAMA ID Enhancement Reduction Dose Benefit Estimated Cost Conclusion ReductionBeet Benefit 033 10.c. Install fuel SAMA would extend 1.39% 2.79% $24,393 $146,356 >$2,000,000 Not cost cells. DC power availability in effective an SBO, which would extend HPCI/RCIC operability and allow more time for AC power recovery.

Basis for

Conclusion:

The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.

034 10.d. Enhance This SAMA would 4.65% 1.91% $19,761 $118,568 $13,000 Retain procedures to improve DC power make use of DC availability.

bus cross-ties.

Basis for

Conclusion:

The CDF contribution due to loss of DC buses D16 and D17 was eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $13,000 by engineering judgment.

E.2-31

3 3 NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 3

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phas IIResut o Potntil CD Of~itUpper Phase II SAMA Result of Potential CDF Dose Estimated Bound Estimated Conclusion SAMA ID Enhancement Reduction Reduction Benefit Estimated Cost RedutionBenefit 035 10.e. Extended SAMA would extend 1.39% 2.79% $24,393 $146,356 $500,000 Not cost SBO provisions. DC power availability in effective an SBO, which would extend HPCI/RCIC operability and allow more time for AC power recovery.

Basis for

Conclusion:

The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

Improvements in Identifying and Mitigating Containment Bypass 036 Locate RHR SAMA would prevent 0.33% 0.21% $2,749 $16,497 >$500,000 Not cost inside ISLOCA outside effective containment. containment. .

Basis for

Conclusion:

RHR ISLOCA accident sequences were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be greater than $500.000. Therefore, this SAMA is not cost effective for PNPS.

037 Increase SAMA could reduce 0.54% 0.38% $4,025 $24,148 $100,000 Not cost frequency of valve ISLOCA frequency. effective leak testing. _ _

Basis for

Conclusion:

The CDF contribution due to ISLOCA was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $100,000 by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

E.2-32

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding AP"-

SAMA Analysis 50-293-LR, 06-848-02-LR

, It_

T_

Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)

Phase II Result of Potential CDF OffSit Estimated Estmaed Upper Bondostmaed Estimated Conclusion SAMA ID Enhancement Reduction Dose ReductionBeet Benefit Estimated Cost 038 8.e. Improve This SAMA would 0.00% 0.00% $0 $0 >$2,000,000 Not cost MSIV design. decrease the likelihood effective of containment bypass scenarios.

Basis for

Conclusion:

Containment bypass failure due to MSIV leakage was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.

Improvements Related to Core Cooling System 039 Install an SAMA would allow 0.00% 0.00% $0 $0 $135,000 Not cost independent continued inventory in effective diesel for the CST CST during an SBO.

makeup pumps.

Basis for

Conclusion:

As currently modeled, if CST water level is low, swapping HPCI/RCIC suction from the CST to the torus allows continued HPCI/RCIC injection. Therefore, the failure to switchover from CST to torus was eliminated to conservatively assess the benefit of this SAMA on CDF. The cost of implementing this SAMA was estimated to be $135,000 by engineering judgment.

Therefore, this SAMA is not cost effective for PNPS.

E.2-33

9 3-NRC - Applicant's Environmental Report SAMA Analysis 3

Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase 11SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase II Result of Potential CDF Offt Estimated Estmaed Upped Estimated Bondostmaed Conclusion SAMA ID S Enhancement Reduction De Benefit Estimated Cost ReductionBefi 040 Provide an SAMA would reduce 3.15% 1.97% $18,369 $110,212 >$2,000,000 Not cost additional high frequency of core melt effective pressure injection from small LOCA and pump with SBO sequences.

independent diesel.

Basis for

Conclusion:

The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.

041 Install SAMA would allow 3.15% 1.97% $18,369 $110,212 >$2,000,000 Not cost independent AC makeup capabilities effective high pressure during transients, small injection system. LOCAs, and SBOs.

Basis for

Conclusion:

The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.

E.2-34

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)

Off-SiteUpper Phase 11 SAMA Result of Potential CDF off-Site Estimated Bound Estimated

Conclusion:

SAMA ID Enhancement Reduction De Benefit' Estimated Cost ReductionBeet 042 2.a. Install a SAMA would improve 3.15% 1.97% $18,369 $110,212 >$2,000,000 Not cost passive high prevention of core melt effective pressure system. sequences by providing additional high pressure capability to remove decay heat through an isolation condenser type system.

Basis for

Conclusion:

The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.

043 2.d. Improved SAMA will improve 2.11% 1.43% $12,671 $76,025 >$2,000,000 Not cost high pressure prevention of core melt effective systems sequences by K improving reliability of high pressure capability to remove decay heat.

Basis for

Conclusion:

The CDF contribution from reducing the HPCI system failure probability by a factor of 3 was estimated to bound the potential impact of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.

E.2-35

Exhibit No. NRC000001 a 3 NRC - Applicant's Environmental Report SAMA Analysis J

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)

Phase 11 Result of Potential CDF OffSt Estimated Bound Estimated SAMA Esimtdoond Esiatd Conclusion SAMA ID Enhancement Reduction ReductionBeet Benefit Estimated Cost 044 2.e. Install an SAMA will improve 3.15% 1.97% $18,369 $110,212 >$2,000,000 Not cost additional active reliability of high- effective high pressure pressure decay heat system. removal by adding an additional system.

Basis for

Conclusion:

The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.

045 8.c. Add a diverse SAMA will improve 3.15% 1.97% $18,369 $110,212 >$2,000,000 Not cost injection system. prevention of core melt effective sequences by providing additional injection capabilities.

Basis for

Conclusion:

The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.

E.2-36

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR It--- 1 It-,

Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase II 1

I Result of Potential CDF OffSiteUpper OffSite Estimated Bound Estimated C SAMA ID SAMA Enhancement Reduction Dose Benefit Estimated Cost I. ~~~~~Reduction Bnft ______

Improvements Related to ATWS Mitigation 046 Increase SRV SAMA addresses the 1.51% 0.92% $10,600 $63,599 $2,000,000 Not cost reseat reliability. risk associated with effective dilution of boron caused by the failure of the SRVs to reseat after SLC injection.

Basis for

Conclusion:

The CDF contribution due to stuck open relief valves was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.

047 11.a. Install an This SAMA would 0.50% 1.19% $10,283 $61,701 >$2,000,000 Not cost ATWS sized vent. provide the ability to effective remove reactor heat from ATWS events.

Basis for

Conclusion:

The CDF contribution from ATWS sequences associated with containment bypass were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing of this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.

E.2-37

Exhibit No. NRC000001 D

NRC - Applicant's Environmental Report SAMA Analysis J J Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase 11 Result of Potential CDF Offit Estimated SAMA Esim tdoo BoundndEsi Estimated atd Conclusion SAMA ID Enhancement Reduction Dos Benefit Estimated Cost ReductionBefi 048 Diversify An alternate means of 0.00% 0.02% $2,153 $12,915 >$200,000 Not cost explosive valve opening a pathway to effective operation. the RPV for SLC system injection would improve the success probability for reactor shutdown.

Basis for

Conclusion:

Common cause failure of SLC explosive valves was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $200,000 by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

Other Improvements 049 Increase the SAMA reduces the 0.73% 0.60% $5,300 $31,799 >$1,500,000 Not cost reliability of SRVs consequences of effective by adding signals medium break LOCAs.

to open them automatically.

Basis for

Conclusion:

The CDF contribution from SRVs failing to open in medium LOCA sequences was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $1.5 million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

E.2-38

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase II Result of Potential CDF Offsit Estimate d Upper Estimated SAMA ID SAMA Enhancement Reduction Dose Reduction:Bnei Benefit Estimated Costnclus 050 8.e. Improve SRV This SAMA would 4.81% 3.51% $32,396 $194,378 >$2,000,000 Not cost design. improve SRV reliability effective thus increasing the likelihood that sequences could be mitigated using low-pressure heat removal.

Basis for

Conclusion:

The probability of SRV failure to open for vessel depressurization was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.

051 Provide self- SAMA would eliminate 0.47% 0.55% $4,902 $29,412 >$200,000 Not cost cooled ECCS ECCS dependency on effective pump seals. the component cooling ,

water system.

Basis for

Conclusion:

The CDF contribution from sequences involving RHR pump failures was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $200,000 by engineering judgment.

Therefore, this SAMA is not cost effective for PNPS E.2-39

Exhibit No. NRC000001 J 3 NRC - Applicant's Environmental Report SAMA Analysis Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)

Phase II SAMA Result of Potential CDF OffSite Estimated Upper Estimated Dose Etmtd B u dE i aed Conclusion SAMA ID Enhancement Reduction Reducton Benefit Estimated Cost Redu tionBenefit 052 Provide digital Upgrade plant 0.07% 0.01% $2,352 $14,109 >$100,000 Not cost large break LOCA instrumentation and effective protection. logic to improve the capability to identify symptoms/precursors of a large break LOCA (a leak before break).

Basis for

Conclusion:

The CDF contribution due to large break LOCA was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $100,000 by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

E.2-40

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phas IIRestto Potntil CF ~ iteUpper PhaseS A ResuMA-of Potential CDF Estimated Bound Estimated o SAMA ID Enhancement Reduction Dose Benefit Estimated Cost. Conclusion ReductionBefi Improvements Related to IPE, IPE Update& IPEEE Insights 053 Control This SAMA would 3.61% 2.24% $22,873 $137,237 $300,000 Not cost containment establish a narrow effective venting within a pressure control band narrow band of to prevent rapid pressure containment depressurization when venting is implemented thus avoiding adverse impact on the low pressure ECCS injection systems taking suction from the torus.

Basis for

Conclusion:

The probability of the operator failing to recognize the need to vent the torus was reduced by a factor of 3 to conservatively assess the benefit of this SAMA on CDF. The cost of implementing this SAMA was estimated to be $300,000 by engineering judgment. Therefore, this SAMA Is not cost effective for PNPS.

E.2-41

J) i NRC - Applicant's Environmental Report SAMA Analysis Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Phase II Result of Potential CDF Estimated Upper Estimated SAMA ID SAMA Enhancement Reduction Dose Benefit Estimated Cost onclus on ReductionBeet 054 Install a bypass This SAMA would 0.28% 0.33% $3,627 $21,761 $1,000,000 Not cost switch to bypass reduce the core effective the low reactor damage frequency pressure contribution from the interlocks of LPCI transients with stuck or core spray open SRVs or LOCAs injection valves cases. Core Spray and LPCI injection valves require a low permissive signal from the same two sensors to open the valves for RPV injection.

Basis for

Conclusion:

The probability of the ECCS low-pressure permissive failing was eliminated to conservatively assess the benefit of this SAMA on CDF. The cost of implementing this SAMA at Dresden was estimated to be $1 million. Therefore, this SAMA is not cost effective for PNPS.

055 Increase the This SAMA would 4.37% 6.63% $59,385 $356,310 >$5 million Not cost reliability of SSW reduce common cause effective and RBCCW dependencies from pumps. SSW and RBCCW systems and thus reduce plant risk.

Basis for

Conclusion:

The CDF contribution from sequences involving common cause failures of SSW and RBCCW was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $5million by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

E.2-42

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)

Phase II SAMA Result of Potential CDF Off-Site Estimated Upper Estimated SAMA ID S Enhancement Reduction Reduction Benefit Estimated Cost RedutionBenefit 056 Provide redundant This SAMA would 8.81% 3.51% $36,773 $220,639 $112,400 Retain DC power improve reliability of supplies to DTV the DTV valves and valves. enhance containment heat removal capability.

Basis for

Conclusion:

The CDF contribution from sequences involving DC power supply failures to the DTV valves was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $112,400 by engineering judgment.

057 Proceduralize use This SAMAwould 2.25% 3.14% $29,213 $175,279 $26,000 Retain of the diesel fire increase capability to pump hydro provide makeup to the turbine in the fire pump- day tank to event of EDG A allow continued failure or operation of the diesel unavailability. fire pump, without dependence on electrical power.

Basis for

Conclusion:

The CDF contribution from sequences involving a LOOP and failure of either EDG A, or the EDG A fuel oil transfer oil pump, was eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be

$26,000 by engineering judgment.

E.2-43

3 3 NRC - Applicant's Environmental Report SAMA Analysis 3

Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)

Upper Phase 11 Result of Potential CDF f-eeneit Buppenr Cst SAMA ID SAMA Enhancement Reduction Dose Benefit Estimated Cost Cnlso 058 Proceduralize the This SAMA would 4.92% 3.14% $31,799 $190,797 $50,000 Retain operator action to provide the direction to feed BI loads via restore B15 and B17 B3 When A5 is loads upon loss of A5 unavailable post- initiating events as long trip. Similarly, as A3 is available.

feed B2 loads via Additionally, it would B4 when A6 is provide the direction to unavailable post restore B14 and B18 trip., loads upon loss of A6 initiating events as long as A4 is available.

Basis for

Conclusion:

The CDF contribution from sequences involving loss of 4160VAC safeguard bus AS was conservatively eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $50,000 by engineering judgment.

059 Provide redundant This SAMA would 8.77% 17.19% $154,966 $929,797 $1,956,000 Not cost path from fire enhance the effective protection pump availability and discharge to LPCI reliability of the loops A and B firewater cross-tie to cross-tie. LPCI loops A and B for reactor vessel injection and drywell spray.

Basis for

Conclusion:

The CDF contribution from sequences involving firewater injection failures was conservatively eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $1,956,000 by engineering judgment.

Therefore, this SAMA isnot cost effective for PNPS E.2-44

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results Upper Upper Upper PhEstimated Bound Estimaited Bound Estimated Bound II Benefit Estimated Benefit Estimated Benefit Estimated SAMASAMA Benefit Cost Benefit Benefit IDBase Line Base Lne Sensitivity Sensitivity Sensitivity Sensitivity Case I Case I Case 2 Case 2 I Install an independent $43,639 $261,832 $5,800,000 $50,320 $301,920 $59,355 $356,129 method of suppression pool cooling.

2 Install a filtered containment $0 $0 $3,000,000 $0 $0 $0 $0 vent to provide fission product scrubbing. Option 1: Gravel Bed Filter Option 2: Multiple Venturi Scrubber 3 Install a containment vent $10,283 $61,701 >$2,000,000 $11,702 $70,211 $14,207 $85,244 large enough to remove ATWS decay heat.

4 Create a large concrete $436,759 $2,620,551 >$100 million $492,136 $2,952,813 $610,307 $3,661,845 crucible with heat removal potential under the basemat to contain molten core debris.

5 Create a water-cooled rubble $436,759 $2,620,551 $19,000,000 $498,057 $2,988,339 $610,307 $3,661,845 bed on the pedestal.

E.2-45

3 J NRC - Applicant's Environmental Report SAMA Analysis

.a..

Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)

Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound II Benefit Estimated Estimated Benefit Estimated Benefit Estimated SAASAMA Benefit CotBenefit Benefit ID Sensitivity Sensitivity Sensitivity Sensitivity BsLie BeLieCase I Case I Case 2 Case 2 6 Provide modification for $2,153 $12,915 >$1,000,000 $2,425 $14,551 $3,008 $18,048 flooding the drywell head 7 Enhance fire protection $10,763 $64,577 >$2,500,000 $12,127 $72,764 $15,040 $90,238 system and/or SGTS hardware and procedures.

8 Create a core melt source $436,759 $2,620,551 >$5,000,000 $498,057 $2,988,339 $610,307 $3,661,845 reduction system.

9 Install a passive containment $44,037 $264,219 $5,800,000 $50,845 $305,069 $59,803 $358,816 spray system.,

10 Strengthen primary/ $205,571 $1,233,428 $12,000,000 $231,636 $1,389,815 $287,257 $1,723,540 secondary containment.

11 Increase the depth of the $4,305 $25,831 >$5,000,000 $4,851 $29,105 $6,016 $36,095 concrete basemat or use an alternative concrete material to ensure melt-through does not occur 12 Provide a reactor vessel $3,229 $19,373 $2,500,000 $3,638 $21,828 $4,512 $27,071 exterior cooling system (see

  1. 7) _

E.2-46

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim I- LR 1 Proceeding A4*1 SAMA Analysis 50-293-LR, 06-848-02-LR 1--

Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)

Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound II Benefit Estimated Benefit Estimated Benefit Estimated SAMA SAMA Benefit Cost Benefit Benefit ID L Sensitivity Sensitivity Sensitivity Sensitivity Base Case I Case I Case 2 Case 2 13 Construct a building to be $10,763 $64,577 >$2,000,000 $12,273 $73,640 $15,040 $90,238 connected to primary/

secondary containment that is maintained at a vacuum 14 2.g. Dedicated Suppression $43,639 $261,832 $5,800,000 $51,067 $306,400 $59,355 $356,129 Pool Cooling 15 3.a. Create a larger volume in $205,571 $1,233,428 $8,000,000 $234,423 $1,406,537 $287,257 $1,723,540 containment.

16 3.b. Increase containment $205,571 $1,233,428 $12,000,000 $234,423 $1,406,537 $287,257 $1,723,540 pressure capability (sufficient pressure to withstand severe accidents).

17 3.c. Install improved vacuum $0 $o >$1,000,000 $0 $0 $0 $0 breakers (redundant valves in each line).

18 3.d. Increase the temperature $2,153 $12,915 $12,000,000 $2,455 $14,728 $3,008 $18,048 margin for seals.

19 5.b/c. install a filtered vent $0 $0 $3,000,000 $0 $0 $0 $0 E.2-47

-3Ad U NRC - Applicant's Environmental Report SAMA Analysis 9,

Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)

Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound 11 Benefit Estimated Estimated Benefit Estimated Benefit Estimated SAMA Benefit Cost Benefit Benefit ID Sensitivity Sensitivity Sensitivity Sensitivity Case I Case I Case 2 Case 2 20 7.a. Provide a method of $2,153 $12,915 >$1,000,000 $2,455 $14,728 $3,008 $18,048 drywell head flooding.

21 13.a. Use alternate method of $10,763 $64,577 >$2,500,000 $12,273 $73,640 $15,040 $90,238 reactor building spray.

22 14.a. Provide a means of $204,495 $1,226,971 $2,500,000 $230,423 $1,382,539 $285,753 $1,714,516 flooding the rubble bed.

23 14.b. Install a reactor cavity $436,759 $2,620,551 $8,750,000 $498,057 $2,988,339 $610,307 $3,661,845 flooding system.

24 Add ribbing to the $205,571 $1,233,428 $12,000,000 $234,423 $1,406,537 $287,257 $1,723,540 containment shell.

25 Provide additional DC battery $24,393 $146,356 $500,000 $27,830 $166,978 $33,598 $201,588 capacity.

26 Use fuel cells instead of lead- $24,393 $146,356 >$2,000,000 $28,207 $169,242 $33,598 $201,588 acid batteries.

27 Modification for Improving DC $19,761 $118,568 $500,000 $23,377 $140,262 $26,044 $156,263 Bus Reliability 28 2.i. Provide 16-hour SBO $24,393 $146,356 $500,000 $28,207 $169,242 $33,598 $201,588 injection.

E.2-48

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)

Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound IP Benefit Estimated Estimated Benefit Estimated Benefit Estimated SSAMA, Benefit Cost Benefit Benefit ID Sensitivity Sensitivity Sensitivity Sensitivity Base Line Base Line Case I - Case 1 Case 2 Case 2 29 9.b. Provide an alternate $44,281 $265,687 >$2,000,000 $50,546 $303,278 $60,956 $365,738 pump power source.

30 9.g. AC Bus Cross-Ties $78,902 $473,410 $146,120 $91,662 $549,972 $106,357 $638,142 31 10.a. Add a dedicated DC $150,504 $903,025 $3,000,000 $178,405 $1,070,432 $201,864 $1,211,183 power supply.

32 10.b. Install additional $150,504 $903,025 $3,000,000 $178,405 $1,070,432 $201,864 $1,211,183 batteries or divisions.

33 10.c. Install fuel cells. $24,393 $146,356 >$2,000,000 $28,207 $169,242 $33,598 $201,588 34 10.d. DC Cross-Ties $19,761 $118,568 $13,000 $23,377 $140,262 $26,044 $156,263 35 10.e. Extended SBO $24,393 $146,356 $500,000 $28,207 $169,242 $33,598 $201,588 provisions.

36 Locate RHR inside $2,749 $16,497 >$500,000 $3,213 $19,276 $3,680 $22,077 containment.

37 Increase frequency of valve $4,025 $24,148 $100,000 $4,688 $28,127 $5,407 $32,444 leak testing.

38 8.e. improve MSIV design. $0 $0 >$2,000,000 $0 $0 $0 $0 E.2-49

3 NRC - Applicant's Environmental Report SAMA Analysis a Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)

Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound Ii Benefit Estimated Benefit Estimated Benefit Estimated SAMA SAMA Benefit Cost Benefit Benefit ID  ; L Sensitivity Sensitivity Sensitivity Sensitivity Base Line Base Line Case I Case I Case 2 Case 2 39 Install an independent diesel $0 $0 $135,000 $0 $0 $0 $0 for the CST makeup pumps.

40 Provide an additional high $18,369 $110,212 >$2,000,000 $21,540 $129,238 $24,477 $146,860 pressure injection pump with independent diesel.

41 Install independent AC high $18,369 $110,212 >$2,000,000 $21,902 $131,415 $24,477 $146,860 pressure injection system.

42 2.a. Install a passive high $18,369 $110,212 >$2,000,000 $21,902 $131,415 $24,477 $146,860 pressure system.

43 2.d. Improved high pressure $12,671 $76,025 >$2,000,000 $14,851 $89,109 $16,894 $101,363 systems 44 2.e. Install an additional $18,369 $110,212 >$2,000,000 $21,902 $131,415 $24,477 $146,860 active high pressure system.

45 8.c. Add a diverse injection $18,369 $110,212 >$2,000,000 $21,902 $131,415 $24,477 $146,860 system.

46 Increase SRV reseat $10,600 $63,599 $2,000,000 $12,326 $73,958 $14,270 $85,623 reliability.

47 11.a. Install an ATWS sized $10,283 $61,701 >$2,000,000 $11,857 $71,142 $14,207 $85,244 vent.

E.2-50

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)

Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound 11 Benefit Estimated Estimated Benefit Estimated Benefit Estimated SSAMA Benefit Cost Benefit Benefit ID Sensitivity Sensitivity Sensitivity Sensitivity Base Line Base Line Case I Case I Case 2 Case 2 48 Diversify explosive valve $2,153 $12,915 >$200,000 $2,425 $14,551 $3,008 $18,048 operation.

49 Increase the reliability of $5,300 $31,799 >$1,500,000 $6,163 $36,978 $7,135 $42,811 SRVs by adding signals to open them automatically.

50 8.e. Improve SRV design. $32,396 $194,378 >$2,000,000 $37,767 $226,602 $43,483 $260,897 51 Provide self-cooled ECCS $4,902 $29,412 >$200,000 $5,638 $33,829 $6,687 $40,125 pump seals.

52 Provide digital large break $2,352 $14,109 >$1 00,000 $2,688 $16,126 $3,232 $19,391 LOCA protection.

53 Control containment venting $22,873 $137,237 $300,000 $26,653 $159,919 $30,716 $184,299 within a narrow band of pressure 54 Install a bypass switch to $3,627 $21,761 $1,000,000 $4,163 $24,978 $4,960 $29,758 bypass the low reactor pressure interlocks of LPCI or core spray injection valves.

55 Improve SSW System and $59,385 $356,310 >$5 million $67,986 $407,918 $81,467 $488,799 RBCCW pump recovery.

E.2-51

J 3 NRC - Applicant's Environmental Report SAMA Analysis

.,)

Exhibit No. NRC000001 Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)

Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound II Benefit Estimated Estimated Benefit Estimated Benefit Estimated SAMA SAMA Benefit Cost Benefit Benefit ID Sensitivity Sensitivity Sensitivity Sensitivity Base Line Base Line Case I Case I Case 2 Case 2 56 Provide redundant DC power $36,773 $220,639 $112,400 $43,541 $261,247 $48,408 $290,449 supplies to DTV valves.

57 Proceduralize the use of $29,213 $175,279 $26,000 $33,568 $201,406 $39,901 $239,406 diesel fire pump hydroturbine in the event of EDG A failure or unavailability.

58 Proceduralize the operator $31,799 $190,797 $50,000 $36,980 $221,878 $42,811 $256,868 action to feed B1loads via B3 When AS is unavailable post-trip.

59 Provide redundant path from $154,966 $929,797 $1,956,000 $176,682 $1,060,091 $213,620 $1,281,720 fire protection pump discharge to LPCI loops A and B cross-tie.

E.2-52

Retrieved from "https://kanterella.com/w/index.php?title=ML110600910&oldid=2121158"