ML110030965

Exhibit NRC000001, Pilgrim Lr Proceeding - Applicant'S Environmental Report, Attachment E, Severe Accident Mitigation Alternatives Analysis.

ML110030965
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: ML110030965 ML110030966 ML110030967 ML110030968 ML110030969 ML110030970 ML110600910
References
RAS 19379, 50-293-LR, ASLBP 06-848-02-LR, NRC000001
Download: ML110030965 (130)

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

Attachment E.1 - Evaluation of PSA Model E.1-Evaluation Model E.2 - Evaluation of SAMASAMA Candidates Candidates

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

Operating License Renewal Stage "-""

Table of Contents EVALUATION OF PROBABILISTIC E.1 EVALUATION PROBABILISTIC SAFETY SAFETY ANALYSIS ANALYSIS MODEL MODEL .... ********* ..... E.1-1 E.1-1 E.1.1 PSA Model - Level 1 Analysis .....................................

E.1.1 PSA ..................................... E.1-1 E.1-1 E.1.2 PSA Model - Level Analysis .....................................

Level 22 Analysis ..................................... E.1-27 E.1-27 E.1.2.1 Containment Containment Performance Analysis .............................

............................. E.1-27 E.1-27 E.1.2.2 ........................

E.1.2.2 Radionuclide Analysis ........*....*..................  :: ...... E.1-33 E.1-33 E.1.2.2.1 E. 1.2.2.1 Introduction ........................................... E.

Introduction .................................... E.1-33 1-33 E.1.2.2.2 Timing of Release ........................

Timing ...................................... E.1-33 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 Progression Bins Source Terms ........... .... ....... E.1-43 E.1-43 E.1.2.2.7 Release Magnitude Calculations ..........................

Calculations .......................... E.1-52 E.1.3 E.1.3 IPEEE ..........................

IPEEE Analysis ................................................. E.1-52 E.1-52 E.1.3.1 Seismic Analysis ..........................

............................................ E.1-52 E.1-52 E.1.3.2 Analysis .......................

E.1.3.2 Fire Analysis ............................................... . . . E.1-52

. E.1-52 I;

~

E.1.3.3 Other External ......................

External Hazards ........*.............................. E.1-54 E.1.4 PSA Model Peer Review and Difference between Current Current PSA Model and 1995 ..

1995 Update IPE ............................................... ..................... E.1-54 E.1.4.1 PSA Model Peer Review Review ........................

...................................... .... E.1 -54 E.1-54 E.1.4.2 Major Differences Differences between the Updated Updated IPE PSA Model and 1995 Update IPE Model .....................................

................. E.1-55 E.1~55 E.1.4.2.1 Core Damage Damage - Comparison Comparison to the PNPS 1995 1995 Update Model .........................

Update IPE Model ........ '......... " .................. E.1-55 E.1-55 E.1.4.2.2 Containment Performance -- Comparison E.1.4.2.2 Containment Comparison to the Original Original .

PNPS IPEIPE Model ..........

...................................... E.1-59 E.1-59 E.1.5 The MACCS2 E.1.5 MACCS2 Model Model - Level 3 3 AnalYSis Analysis ...............................

. ............................ E.1-60 E.1-60 Introduction ................................................

E.1.5.1 Introduction ................. I........ E.1-60 E.1-60 E.1.5.2 Input .....................................................

E.1.5.2 Input ................ E.1-60 E.1-60 E.1.5.2.1 Projected Total Population by Spatial Element ............... ..... .......... E.1-61 E.1.5.2.2 Land E.1.5.2.2 Fraction ....................

Land Fraction ......................................... ........ E.1-62 E.1-62 E.1.5.2.3 E.1.5.2.3 Watershed Class Class ..... ...................................... E.1-62 E.1-62 i ( 1

"'-'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 Applicant's Environmental ili ibl Operating Operating License Renewal Renewal Stage E.1.5.2.4 Regional Economic Data E.1.S.2.4 Data -. ............ ...................

~" ................................ E.1-62 .

E.1.5.2.5 Agriculture Data E.1.S.2.S .................... ...................

Data ....................................... E.1-63 .

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

E.1.S.2.6 ...................

, .................... E.1-63 .

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

E.1.S.2.7 Emergency .

E.1.5.2.8 Core Inventory ........................................

E.1.S.2.B ..................... 1-64 E.1-64 .

E.1.5.2.9 Source Terms E.1.S.2.9 Terms ........................

....................... , ............ '; ..' .. 'E.1-66 E.1-66 .

E.1.5.3 E.1.S.3 Results .....................

................................ ,., .... , .....................*. E.1-66 .

i E.1.6 ............................... , ..... ~ ............... ~ ....*.... E.1-69 E.1.6 References .............*...... E.1-69 .

EVALUATION OF SAMA CANDIDATES E.2 EVALUATION CANDIDATES .......... ................... ....................

........... E.2-1 E.2-1 Compilation ..........................................

E.2.1 SAMA List Compilation ........................................... E.2-1 E.2-1 E.2.2 Qualitative Screening of SAMA ...... E.2 ............. E.2-2*

SAMA Candidates (Phase I) ...................

E.2.3 Final Screening and Cost Benefit Benefit Evaluation of SAMA SAMA Candidates (Phase II) E.2-2 E.2-2 E.2.4 Sensitivity Analyses .............................................

............................................ E.2-11 E.2-11 E.2.5 References .......................

E.2.S .........................

'..... '........................ E.2-13 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 Applicant's Environmental Report Environmental Report '"',,

Operating License license Renewal Stage ~'Ij List of Tables E.1-1 Table E.1-1 Core Damage Damage Frequency Uncertainty.......................................

..............................*........E.1-2 E.1-2 E.1-2 Table E.1-2 Model CDF PNPS PSA Model CDF Results by Major Initiators ........

Major Initiators ..... " ......................E.1-3 E.1-3 E.1-3 Table E.1-3 Correlation of Level 11 Risk Significant Terms to Evaluated SAMAs ................

SAMAs ................ E.1-4 E.1-4 Table E.1-4 Summary of PNPS PSA Core Damage Summary Damage Accident Class .........................

.................. ,. ...... E.1-28 E.1-28 E.1-5 Table E.1-5 Notation and Definitions for PNPS CET CET Functional Nodes Description Description .............

............. E.1-29 E.1-29 Table E.1-7' E.1-7 PNPS Release Categories Categories .....................

..................* , *........................*.E.1-35 Table E.1-6 E.1-6 Release Severity and Timing Classification Scheme Scheme SummarySummary ...................

................... E.1-35 E.1-35 E.1-8 Table E.1-8 Summary of PNPS Core Damage Summary Damage Accident Sequences Plant Damage Damage States ....... E.1-36 States ....... E.1-36 E.1-9 Table E.1-9 Collapsed Accident Progression Bins (CAPB) (CAPB) Descriptions ......................

......................E.1 -44 E.1-44 E.1-10 Table E.1-10 Summary of PNPS Containment Summary Containment Event Tree Quantification ......................

......................E.1-49 E.1-49 E.1-11 Table E.1-11 Collapsed Accident Progression Progression Bin (CAPB) .....................

(CAPB) Source Terms ..................... E.1-50 E.1-50 Table E.1-11 Collapsed Accident Progression Bin (CAPB) (CAPB) Source Terms (continued) ............................................................

...........................................................E. 1-51 E.1-51 E.1-12 Table E.1-12 PNPS Fire Updated Updated Core Core Damage Damage Frequency Frequency Results .........................

.........................E.1-53 E.1-53 E.1-13 Table E.1-13 Estimated Population Distribution within a 50-mile Radius .......................

Estimated .......................E.1-61 iii iii &:

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Nuclear Power Pilgrim Nuclear Pilgrim PowerStation Station Applicant's Environmental Report Applicant's Environmental Report Operating License Operating Renewal Stage license Renewal Stage Table E.1-14 Table E.1-14 PNPS Core Inventory PNPS Core ........................................E.1-65 (Becquerels) ........................................

Inventory (Becquerels) E.1-65 Table E.1-15 Table E.1-15 Base Case Base Mean PDR Case Mean PDR and OECR Values and OECR .......................... '" ......E.1-67 Values................................... E.1-67 Table E.1-16 Table E.1-16 Summary of Summary Offsite Consequence of Offsite Sensitivity Results Consequence Sensitivity ..........................E.1-68 Results .......................... E.1-68 Table Table E.2-1 E.2-1 Summary of Summary Phase II of Phase SAMA Candidates II SAMA ConSidered in Candidates Considered in Cost-Benefit Evaluation .....

Cost-Benefit Evaluation .....E.2-15 E.2-15 Table Table E.2-2 E.2-2 Sensitivity Analysis Results..............................................E.245 Sensitivity Analysis Results ...............................................E.2-45 iv iv

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

List list of Figures Figure E.1-1 Figure E.1-1 PNPS Radionuclide Release Category Summary Release Category Summary .........................

......................... E.1-31 Figure Figure E.1-2 E.1-2 Contribution to LERF Damage State Contribution PNPS Plant Damage ........... .............. E.1-32 LERF .........................

v v0

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

,f ".

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Power Station Applicant's Environmental Report Applicant's Environmental Report Operating License Renewal Stage Ui E.1 EVALUATION OF PROBABILISTIC EVALUATION PROBABILISTIC SAFETY SAFETY ANALYSIS MODEL MODEL The severe accident risk was estimated using the Probabilistic Safety Analysis (PSA) (PSA) model model and a Level 3 model developed using the MACCS2 MACCS2 code. The CAFTA CAFTA code was used to develop the Pilgrim Nuclear Power Station (PNPS) (PNPS) PSA Level 1 I and Level 2 models. This section provides the description description of PNPS PSA Levels 1, 1, 2, and 3 analyses, Core Damage Damage Frequency (CDF) (CD F) uncertainty, Individual Plant Examination Examination of External External Events (IPEEE)

(IPEEE) analyses, and PSA model model peer review.

E.1.1 PSA Model-Model - Level I1 Analysis The PSA model (Level 1 I and Level 2) used for the SAMA SAMA analysis was the most recent internal internal events risk model for PNPS (Revision 1, 1, April 2003) 2003) [Reference E.1-1]. The PNPS PSA model model and documentation documentation has been updated updated to reflect the current plant operating configuration and design changes as of September September 2001. The current PSA model model reflects the accumulation of additional plant operating history and component failure and unavailability data as of December December 2001. The PSA model also resolves all findings and observations during the industry peer review of the model, conducted in March March 2000 [Reference E.1-1]. The PNPS model adopts the tree/ large fault tree approach and uses the CAFTA small event tree/large CAFTA code for quantifying CDF. CDF. The Level I1 and Level 2 PNPS PSA analyses were originally developed and submitted to the NRC in September 19921992 as the Pilgrim Nuclear Power Station Individual Plant ExaminationExamination (IPE) (IPE)

Submittal [Reference E.1-2].

[Reference E.1-2].

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

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

performance.

• • Equipment performance - As data collection progresses, estimated failure rates and system unavailability data change. .

• • Plant configuration changes - Plant configuration changes are incorporatedincorporated into the PSA PSA model.

model.

• • Modeling changes - The PSA model is refined to incorporate the latest state of knowledge and recommendations recommendations from internal and industry peer reviews.

The PSA model contains the major initiators leading to core damage with baseline CDFs CDFs listed inin 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. CDF-based CDF-based Risk Reduction Worth (RRW) (RRW) rankings were reviewed down to 1.005. Events below this point would influence the CDF CDF by less than 0.5% and E.1-1 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 cost-beneficial enhancements. These basic events, including component failures, operator actions, and initiating enhancements.

events, were reviewed to determine if additional SAMA SAMA actions 'may may need to be considered.

Table E.1-3 provides a correlation between the Level 11 RRW risk significant events (component (component failures, failures"operator events) down to 1.005 operator actions, and initiating events) 1.005 identified from the PNPS PSA PSA SAMAs evaluated in Section E.2.

model and the SAMAs The uncertainty associated with CDF CDF was estimated using Monte Carlo techniques implemented implemented CAFTA for the base case mode.

in CAFTA mode. The results are shown in Table E.1-1.

Table E.1-1

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

CDF (\RY)

Mean value 6.68E-6 6.68E-6 5 th perce Sth percentile.30E-6 4.30E-6 l1tile SOth 5 0 th percentile S.93E-6 5.93E-6 th '1.08E-S 1.08E-5 v 95th percentile 9S E.1-1 reflect the uncertainties associated with the data distributions used in The values in Table E.1-1 in th the analysis. The ratio of the 95 9 5 th percentile to the mean is about 1.62. This uncertainty factor is is included in the factor of 6 used to determine the upper bound estimated benefit described in in Appendix E,E, Section 4.21.5.4.

E.1-2 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 Environmental Report Report ":

Operating License license Renewal Stage ~

E.1-2 Table E.1-2 CDF Results by Major Initiators PNPS PSA Model CDFResults CDF Percentage of HE Type IE IE Description CDF Percentage of (fRY) ,CDF TDC TOC Loss of DC DC Power Buses 3.06E-06 3.06E-06 47.77%

LOOP Loss of Offsite Power 1.29E-06 20.12%

TAC Loss of AC Power Buses 8.83E-07 B.B3E-07 13.78%

13.7B%

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

6.10%

TRAN Transients 3.60E-07 5.62%

5.62%

LOCA LOCA Loss of Coolant Coolant Accident

, 1.75E-07 1.75E-07 2.73%

2.73%

SBO Station Blackout 1.46E-07 1.46E-07 2.28%

2.2B%

ATWS ATWS Transient Without Anticipated Transient Without Scram Scram 5.30E-08 5.30E-OB 0.B3%

0.83%

ISLOCA IS LOCA Interfacing System LOCA Interfacing 3.64E-OB 3.64E-08 0.57%

0.57%

FLOOD FLOOD Internal Flooding 1.28E-08 1.2BE-OB 0.20%

0.20%

Total 6.41E-06 100.00%

100.00%

E.1-3 E.1-3 . u

Exhibit No. NRC000001 AC" NRC - Applicant's Environmental Report SAMA Analysis r

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Environmental Report OperatinQ Operating License Renewal Stage Table E.1-3 Table E.1-3 Correlation Correlation of Level 1 I Risk Significant Terms Terms to Evaluated Evaluated SAMAs Event Event Name Name Probability Probability RRW RRW Event Description Event Description Disposition IE-T1 IE-T1 6.70E-02 1.337 1.337 Loss of Loss of offsite offsite This term represents represents the the LOOP LOOP initiating event.

event. Industry Industry efforts efforts power (LOOP) power (LOOP) over the last twenty years have led over led to aa significant Significant reduction in in plant scrams plant scrams from all causes. Improvements Improvements relatedrelated toto enhancing offsite power availability or reliability enhancing reliability and and coping coping with with SBO events were already SSO already implemented implemented and evaluatedevaluated during Phase II SAMA SAMA screening. Phase II II SAMAs SAMAs 025, 026, 027, 028, 028, 029, 030, 030, 033, 033, and and 035 for for enhancing enhancing AC or or DC system reliability DC system reliability or to cope with LOOP and or and SBO events were SSO events were evaluated.

IE-TDCB IE-TOCS 2.63E-03 2.63E-03 1.319 1.319 Transient caused This term represents represents anan initiating initiating event caused by aa complete by loss by loss of 125VOC 125VDC loss of 125VOC l25VDC buses D-17,0-17, D5, 05, and D37 037 and random failures of of bus B busS D-2. Phase II SAMAs battery 0-2. SAMAs to improve battery charging capability and replace replace existing batteries batteries with with more reliable reliable ones ones have already have already been been installed. Phase Phase IIIISAMAs SAMAs 025, 025,026, 027, 031, 026, 027, 032, 033, 034, and 035 035 for enhancing enhancing DC DC system system availability and reliability were reliability were evaluated.

evaluated.

IE-TDCA IE-TOCA 2.63E-03 2.63E-03 1.304 1.304 Transient caused Transient This term represents an initiating event caused by a a complete by loss of 125VOC 125VDC 125VDC buses 0-16, loss of 125VOC D-16, 04, D4, and D36,036, and random random failures of bus A D-1. Phase I SAMAs battery 0-1. SAMAs to improve battery charging

-- ~ c_' __ -

capability and capability and replace existing existing batteries batteries with moremore reliable ones have already been installed.

have installed. Phase Phase IIIISAMAs SAMAs 025, 026, 026, 027, 031, 031, 032, 033, 032, 033, 034, 034, and and 035 for enhancing DC DC system availability availability and reliability were evaluated.

E.1-4 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 Environmental Report Applicant's Environmental Operating License Renewal Stage Operating Table E.1-3 E.1-3 Correlation I Risk Significant Terms Correlation of Level 1 Terms to Evaluated Evaluated SAMAs Name Event Name Probability RRW Event Description Event Description Disposition Disposition FXT-XHE-FO-V4T2 FXT-XHE-FO-V4T2 2.31 E-02 2.31E-02 1.121 1.121 Operator fails to Operator fails This term to align failure to represents operator failure term represents align fire water via fire water via the the.

align fire water LPCI injection path for alternate RPV vessel injection. Phase I crosstie for reactor SAMAs, improvement of procedures and installation of SAMAs, including improvement of pressure vessel instrumentation to enhance the likelihood of success of operator operator (RPV)

(RPV) injection via in response to accident conditions, have already been action in LPCI (transient)

(transient) Phase II SAMAs implemented. Phase" SAMAs 057 and 059, which 057 and which recommend recommend proceduralizing use of the diesel fire pump hydroturbine following proceduralizing EDG A failure, and providing a redundant path from fire water pump discharge to LPCI loops A and B LPClloops B cross-tie, were evaluated.

AC2-PHN-PE-23kV AC2-PHN-PE-23kV 5.OOE-01 5.00E-01 1.079 1.079 Loss of Loss of shutdown shutdown represents loss of term represents This term transformer 23kV of the shutdown transformer 23kV feed feed transformer 23kV transformer 23kV to 4.16kV bus A8. Improvements Improvements related to enhancing offsite feed power availability or reliability and coping with SBO SSO events were already implemented implemented and evaluated during during Phase II SAMASAMA Phase II SAMAs screening. Phase" SAMAs 025, 026, 027, 028, 029, 030, 033, 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 IE-TSSW 6.85E-05 1.065 1.065 Loss of salt service service This term represents an initiating event caused by a a complete water (SSW)

(SSW) loss of the service water system. Phase I SAMAs were system implemented to improve service water system reliability by implemented by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall wall between the SSW pumps. Phase IIII SAMA SAMA 055 to improve SSW SSW system reliability by reducing common dependencies was evaluated.

E.1-5 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 Environmental Report Report Operating License Renewal StageStage E.1-3 Table E.1-3 of Level 1 Correlation of I Risk Significant Terms to EvaluatedEvaluated SAMAs Name Event Name Probability RRW Event Description . (

Disposition IE-TAC6 IE-TAC6 2.63E-03 1.059 Transient caused This term This term represents loss loss of 4.16kV 4.16kV bus A6.

A6. Phase Phase II SAMAs SAMAs to to by loss of 4160VAC 4160VAC improve 4.16kV bus cross-tie capability and revise procedures procedures to bus busA6 A6 repair or replace failed 4.16kV 4.16kV breakers have already been been implemented.

implemented. PhasePhase IIII SAMAs SAMAs 025,026,027,028,029,030, 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC DC system reliability or to cope with LOOP LOOP and SBO SBO events events were evaluated.

~

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

implemented. Phase Phase II SAMA 053 to control containment venting within a narrow SAMA narrow pressure band to prevent rapid containment depressurization depressurization during venting was evaluated.

IE-TAC5 IE-TAC5 2.63E-03 2.63E-03 1.052 Transient caused This term represents an initiating event caused by loss of 4.16kV 4.16kV by loss of 4160VAC 4160VAC bus AS.

A5. Phase I SAMAs SAMAs to improve 4.16kV bus cross-tie bus busA5AS capability and revise procedures procedures to repair repair or replace failed 4.16kV 4.16kV breakers have already been implemented.

implemented. Phase II II SAMAs SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC 026,027,028,029,030,033, DC system reliability or to cope with LOOP and SBO seo events were evaluated.

E.1-6 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 Environmental Report Operating License Renewal Stage Stage Table Table E.1-3 E.1-3 Correlation of Level 1I Risk Significant Terms Evaluated SAMAs Terms to Evaluated Name Event Name Probability Probability RRW RRW Event Description Event Description Disposition RHR-MAI-MA-HTXAP RHR-MAI-MA-HTXAP 4.08E-04 4.0BE-04 1.051 1.051 RHR heat RHRheat This term represents RHR This RHR heat exchanger E-207A exchanger E-207 A unavailable unavailable exchanger E-207A E-207A due to maintenance, maintenance, leading to loop A RHR suppression pool unavailable due to cooling and drywell spray modes being unavailable for maintenance maintenance containment pressure reduction. Phase II SAMAs containment SAMAs have alreadyalready implemented to use firewater for drywell spray and to use been implemented venting via DTV path to reduce containment pressure. Phase II II SAMAs SAMAs 001, 009, 014, and 059, 059, to provide 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 RBC-MAI-MA-LOOPA 3.71 E-04 3.71E-04 1.046 1.046 RBCCW loop RBCCW loop A A out out This This term represents RBCCW term represents RBCCW loop loop A A unavailable unavailable duedue toto for maintenance maintenance. A Phase I SAMA SAMA was implemented to improve RBCCW system reliability by making component cooling water RBCCW SAMA 055 to Improve trains separate. Phase IIII SAMA improve RBCCW RBCCW system reliability by reducing common dependencies was evaluated.

FXT-XHE-FO-DWS FXT-XHE-FO-DWS 2.21 E-02 2.21E-02 1.046 1.046 Operator fails to Operator This term represents operator failure to align fire water via term represents via the align fire water LPCI injection path for alternate drywell spray to remove cross-tie for for drywell containment heat. Phase II SAMAs, including improvement improvement of spray procedures and installation of instrumentation to enhance the likelihood of success of operator action in in response to accident conditions, have already been implemented.

implemented. Phase II II SAMAs 057 and 059, which recommend recommend proceduralizing use of the diesel diesel fire pump hydroturbine hydroturbine following EDG A failure, and providing aa redundant path from fire water pump discharge to LPCI LPCI loops A A and B B cross-tie, were evaluated.

E.1-7 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 Environmental Report Applicant's Environmental Operating License Renewal Stage E.1-3 Table E.1-3 Correlation of Level 1 I Risk Significant Terms to Evaluated SAMAs SAMAs Name Event Name Probability RRW RRW Event Event Description Disposition Disposition AC8-CBR-CO-204 AC8-CBR-CO-204 9.50E-05 9.50E-05 1.044 1.044 480V 480V circuit breaker This term represents random failure of 480V circuit breaker 52-52-204 fails to 204, leading 204, leading to loss ofof power power to to 480V motor control control center center (MCC)

(MCC) remain closed remain closed B14 andand its its associated loads. A A Phase Phase II SAMA SAMAwas was implemented implemented to proceduralize proceduralize operator operator action action to to manually close close the the circuit circuit breaker. Phase breaker. Phase IIII SAMAs SAMAs 030 and 058 058 to to improve improve 480V bus bus availability were evaluated.

AC8-CBR-CO-103 AC8-CBR-CO-103 9.50E-05 1.044 480V circuit breaker This term represents random random failure of 480V circuit breaker 52-fails to 52-103 fails 103, 103, leading to loss of power to 480V MCC MCC B B15 15 and its remain closed closed associated loads. A Phase I SAMA SAMA was implemented to proceduralize operator proceduralize operator action action to manually manually close thethe circuit circuit breaker. Phase IIII SAMAs 030 and 058 to improve 480V bus availability were evaluated.

FXT-ENG-FR-P140 FXT-ENG-FR-P140 1.92E-02 1.92E-02 1.043 1;043 Diesel fire pump P- P- This term represents diesel fire pump P-140 failure to run. Phase 140 fails fails to run run "11 SAMA 045, to SAMA to add aa diverse injection injection system and provide an an injection source other injection other than fire water, was evaluated.

LCI-HTX-VF-E207AA LCI-HTX-VF-E207 3.24E-04 3.24E-04 1.04 Loop B B heat This term term represents random failure of RHR heat exchanger E- E-E-207A exchanger E-207 A 207A, 207A, leading to loop A RHR suppression pool cooling and failure failure drywell drywell spray modes being unavailable for containment pressure reduction. Phase reduction. Phase II SAMAs SAMAs have already already been implemented implemented to to use firewater for drywell spray and to use venting via DTV path to reduce containment pressure. Phase IIII SAMAs 001, 009, 014, 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 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 Nuclear Power Station Applicant's Environmental Environmental Report Operating License Renewal Stage Stage Table E.1-3 E.1-3 Correlation of Level I1 Risk Significant Terms to Evaluated SAMAs SAMAs Event Name Probability RRW Event Description Event Description Disposition Disposition LCI-HTX-VF-E207B LCI-HTX-VF-E207B 3.24E-04 1.039 1.039 Loop A heat Loop This term represents random This failure of random failure of RHR heat heat exchanger exchanger E- E-exchanger E-207B 2078, leading to loop B 207B, B RHR suppression suppression pool COOling cooling and failure drywell spray modes being unavailable for containment pressure reduction. Phase I SAMAs implemented to SAMAs have already been implemented use firewater for drywell spray and to use venting via DTV DTV path to reduce containment pressure. Phase II 1/ SAMAs SAMAs 001, 001, 009, 014, and 059, to provide alternate means of suppression pool cooling and drywell spray and to enhance the availability and reliability of of firewater for reactor vessel injection and drywell spray, were evaluated.

IE-T2 8.90E-02 1.038 1.038 Loss ofPCS 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 1/ SAMA consequences of this event, was evaluated.

RHR-MAI-MA-HTXBP RHR-MAI-MA-HTXBP 2.69E-04 1.032 RHR heat RHRheat This term represents RHR heat exchanger E-207B E-207B unavailable exchanger E-207B . due to maintenance, leading to loop B B RHR suppression pool pool unavailable due to cooling and drywell spray modes being unavailable for maintenance containment pressure reduction. Phase IJ SAMAs SAMAs have already already been implemented implemented to use firewater for drywell spray and to use venting via DTV DTV path to reduce containment containment pressure. Phase 1/ II SAMAs SAMAs 001, 009, 014, and 059, to provide alternate means of 001,009,014, 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 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 Power Station Applicant's Environmental Environmental Report Operating License Renewal Stage E.1-3 Table E.1-3 Correlation of Level 1I Risk Significant Terms to Evaluated SAMAs Correlation Event Event Name Name Probability RRW Event Description Disposition RBC-MAI-MA-LOOPB RBC-MAI-MA-lOOPB 2.36E-04 2.36E-04 1.029 RBCCW loop Bout RBCCW B out This term represents RBCCW RBCCW loop B unavailable due to for maintenance maintenance. A Phase I SAMA maintenance. SAMA was implemented implemented to Improve improve RBCCW system reliability by making component cooling water trains separate. Phase IIII SAMA 055 to improveimprove RBCCW RBCCW system reliability by reducing common dependencies was evaluated.

DWS-XHE-FO-W2 DWS-XHE-FO-W2 2.85E-04 1.026

.1.026 Operator fails to Operator This term represents operator operator failure to align the drywell sprayspray align drywell spray mode of RHR for containment containment pressure reduction. Phase I mode of RHR SAMAs, SAMAs, including improvement of procedures procedures and installation of of instrumentation to enhance the likelihood instrumentation likelihood of of success of operator operator

~ - ~

action in response to accident conditions, have already been been

~ ~ ~

~ - implemented. No additional Phase 1I implemented. II SAMAs were recommended recommended for this Subject.

subject.

SPC-XHE-FO-WI SPC-XHE-FO-W1 1.54E-04 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 containment pressure pressure reduction.

pool cooling mode Phase I SAMAs, including improvement improvement of procedures procedures and of RHR ofRHR installation of instrumentation instrumentation to enhance the likelihood of I

success of operator operator action in response to accident conditions, ~

have already been implemented.

implemented. No additional Phase IIII SAMAs were recommended recommended for this subject.

LCS-CCF-PG-STNRS lCS-CCF-PG-STNRS 2.22E-05 2.22E-05 1.024 1.024 Common Common cause This term term represents common common cause failure of the core core spray and failure of strainers RHR suction strainers.

strainers. A Phase I SAMA, SAMA, installing improved improved BS~8002A&B BS-8002A&B passive emergency emergency core cooling system system (ECCS)

(ECCS) suction plugged ~ strainers, has been implemented.

implemented. Phase Phase"II SAMAs SAMAs 042,042, 044, and and 045, which recommend addition of independent Independent injection systems to mitigate mitigate this failure event, were evaluated.

E.1-10 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 Environmental Report Applicant's Environmental Operating License Renewal Stage Table E.1-3 Table E.1-l Correlation of Level 1I Risk Significant Terms Terms to Evaluated Evaluated SAMAs Event Name Event Name Probability Probability RRW RRW Event Description Event Disposition DC1-CBR-CO-7216A DC1-CBR-CO-7216A 5.11E-05 1.023 1.023 125VDC 125VDC circuit This tenn term represents represents random failure failure of of 125VDC circuit breaker 125VDC circuit breaker

, breaker 72-16A 72-16A 72-16A, leading to loss of DC power to bus 0-16. D-16. Phase II fails to remain SAMAs SAMAs to improve battery charging capability and replace closed closed existing batteries with more reliable ones have already been installed. Phase IIII SAMAs SAMAs 025, 026, 027, 031, 032, 033, 034, 025,026,027,031,032,033,034, and 035 for enhancing DC system availability and reliability were evaluated.

ADS-XHE-FO-XlT2 ADS-XHE-FO-X1T2 6.88E-04 6.88E-,04 1.023 1.023 Operator fails Operator fails to This tenn term represents represents operator failure failure to manually manually open the SRVs SRVs emergency perform emergency for depressurization depressurization during transients. Phase II SAMAs, SAMAs, including depressurization depressurization improvement of procedures and installation of instrumentation improvement instrumentation to (transient)

(transient) enhance the likelihood of success of operator operator action in in response to accident conditions, have already been implemented.

implemented. No No additional Phase IIII SAMAs SAMAs were recommended recommended for this subject.

DCl-CBR-CO-72165 DC1-CBR-CO-72165 ' 5.11E-05 5.11E-05 1.023 1.023 125VDC circuit 125VDC This tenn term represents random failure of DC circuit breaker 72-165 breaker 72-165 fails to provide power to DTVDTV valve AO 5025, causing failure of the to remain closed valve to open on demand, resulting in in loss of containment venting capability. Phase IIII SAMA SAMA 056 to improve DTV valve availability was evaluated.

OSP-SBO OSP-SBO 7.64E-02

,7.64E-02 1.023 1.023 Operator fails to 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)

(SBO) SAMAs, including improvement SAMAs, procedures and training improvement of SBO procedures diesel to either bus to enhance the likelihood of success of operator action in AS or A6 A50rA6 response to accident conditions, have already already been implemented.

been implemented.

No additional Phase II II SAMAs SAMAs were recommended 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 Power Station Station Environmental Report Applicant's Environmental Operating License Renewal Stage Table E.1-3 E.1-3 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs SAMAs Event Name Probability RRW Event Description Disposition DCI-CBR-CO-7217A DC1-CBR-CO-7217A 5.11E-05 5.11E-05 1.023 125VDC circuit 125VDC This term represents random failure of 125VDC 125VDC circuit breaker breaker breaker 72-17A 72-17A, 72-17 A, leading to loss of DC power to bus D-17.

0-17. Phase II fails to remain remain SAMAs to improve battery charging capability and replace SAMAs closed existing batteries with more reliable ones have already been been Phase II SAMAs installed. Phase" SAMAs 025, 026, 027, 031,031, 032, 033, 034, and 035 for enhancing DC system availability and reliability were evaluated.

OSP-14 OSP-14 4.10E-02 4.10E-02 1.022 1.022 Failure to recover This term represents operator failure to recover offsite power 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 a LOOP event. Phase II SAMAs, 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 improvement procedures and training to enhance the likelihood of success of operator action in in response to accident conditions, have already been implemented.

implemented. No additional Phase 11 SAMAs were recommended for this subject.

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

Phase II \I SAMA SAMA 038 to improve MSIV MSIV design and mitigate mitigate the

. consequences of this event was evaluated.

evaluated .

FXT-MAI-MA-P140 FXT-MAI-MA-P140 9.22E-03 9.22E-03 1.019

. 1.019. Diesel driven fire This term term represents diesel fire pump P-140 in in maintenance.

maintenance.

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

. . .... - . . maintenance E.1-12 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 Pilgrim Nuclear Nuclear Power Power Station Station Applicant's Environmental Report Applicant's Environmental Report Operating License Renewal Stage Operating Stage Table Table E.1-3 E.1-3 Correlation of Level I1 Risk Significant Terms Correlation Evaluated SAMAs Terms to Evaluated Event Name Event Name Probability RRW RRW Event Description Event Disposition AC4-RCK-NO-604 AC4-RCK-NO-604 2.51 E-03 2.51E-03 1.019 1.019 4.16kV circuit This term represents control circuit of 4.16kV represents failure of the control circuit 4.16kV circuit breaker 152-604 breaker 152-604 breaker 152-604, leading to LOOP LOOP to safety bus A6. A6. Phase I control circuit no SAMAs 4.16kV bus cross-tie capability and revise SAMAs to improve 4.16kV output procedure to repair or procedure replace failed 4.16kV or replace 4.16kV breakers breakers have already been installed. In already In addition, a Phase I SAMA SAMA was implemented to proceduralize operator implemented manually close operator action to manually breaker. Phase II the circuit breaker. II SAMAs SAMAs 025, 026, 027, 028, 029, 025,026,027,028,029, enhancing AC or DC system reliability or to 030, 033, and 035 for enhancing cope with LOOP and SBO events were evaluated.

DC1-CBR-CO-72175 OC1-CBR-CO-72175 5.11E-05 5.11E-05 1.018 125VDC 125VOC circuit This term represents random failure of breaker 72-175 of DC circuit breaker 72-175 breaker 72-175 fails to provide power to OTV 5042B, causing failure *of DTV valve AO 5042B, of the to remain closed valve to open on demand, resulting in containment venting in loss of containment SAMA 056 to improve DTV capability. Phase IIII SAMA OTV valve availability was evaluated.

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

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

containment improve OTVDTV valve availability was evaluated.

E.1-13

Exhibit No. NRC000001 SAMA Analysis-J~-

r NRC - Applicant's*10--

Environmental Report Pilgrim LR Proceeding

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

- -""..-Vic - IL, z Pilgrim Pilgrim Nuclear Power Station Applicant's Environmental Environmental Report Operating Operating License Renewal Stage Table E.1-3E.1-3.

Correlation of Level 1I Risk Significant Terms Terms to Evaluated SAMAs SAMAs Event Name Name Probability RRW Event Description Disposition Disposition AC4-RCK-NO-504 AC4-RCK-NO-504 2.51 E-03 2.51E-03 1.017 1.017 4.16kV circuit This term temi represents represents failure of the control circuit of 4.16kV circuit breaker 152-504 breaker breaker 152-504, 152-504, leading to lOOPLOOP to safety bus A5. Phase II control circuit no SAMAs to improve improve 4.16kV bus cross-tie capability and revise output output procedures procedures te>to repair or replace failed 4.16kV breakers have already been installed. Inaddition, a Phase II SAMAwas installed .. In SAMA was implemented to proceduralize operator action to manually implemented manually close the circuit breaker. Phase" Phase II SAMAs SAMAs 025, 026, 027, 028, 029, 030, 033, 030, 033, and 035 035 for for enhancing enhancing AC AC or DC system system reliability or to to cope with LOOP lOOP and SBO SSO events were evaluated evaluated..

SSW-MDP-FS-P208D SSW-MDP-FS-P208D 2.022-03 2.02E-03 1.017 SSW pump P-208D SSW This term represents random This random failure failure of SSW SSW pump pump P-208D P-208D toto fails to start on start. Phase II SAMAs SAMAs were implemented implemented to improve service demand water system reliability by enhancing screen wash, adding redundant redundant DC control powerpower for SSW pumps, pumps, and increasing

, seismic integrity of the partition wall between the SSW pumps.

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

SSW-CCF-FS-3P208 SSW-CCF':FS-3P208 2.26E-05 2.26E-05 1.017 1.017 Common Common cause This term represents common cause failure of 3 service water water failure of 3 SSW SSW pumps to start. Phase I SAMAs SAMAs were implemented implemented to improve pumps to start 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 SSW pumps. Phase Phase"II SAMA SAMA 055 to improve improve SSW system reliability by by reducing common dependencies was evaluated.

E.1-14 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 Nuclear Power Pilgrim Nuclear Pilgrim Power Station Station Applicant's Applicant's Environmental Environmental Report Report Operating Operating License License Renewal Renewal Stage Stage Table Table E.1-3 E.1-3 Correlation of Level Correlation Level*1I Risk Terms to Significant Terms Risk Significant Evaluated SAMAs to Evaluated SAMAs Event Name Event Name Probability Probability RRW RRW Event Description Event Description Disposition Disposition SSW-MDP-FS-P208E SSW-MDP-FS-P208E 2.02E-03 2.02E-03 1.016 1.016 SSW pump SSW pump P-208E P-208E This term This represents random term represents failure of random failure SSW pump of SSW P-208E to pump P-208E to fails fails to start on to start on start.

start. Phase SAMAs were Phase II SAMAs implemented to were implemented to improve improve service service demand demand reliability by system reliability water system water enhancing screen by enhancing wash, adding screen wash, adding redundant DC redundant power for control power DC control pumps, and SSW pumps, for SSW and increasing increasing

- seismic integrity of seismic integrity of the the partition between the wall between partition wall SSW pumps.

the SSW pumps.

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

evaluated.

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

from medium LOCA.

LCS-STR-PG-8002A LCS-STR-PG-8002A 1.20E-04 1.20E-04 1.014 1.014 ECCS strainer BS-ECCS strainer BS- This failure of represents failure term represents This term of core spray and core spray RHR suction and RHR suction 8002A plugged 8002A plugged BS-8002A. A strainer BS-8002A.

strainer A Phase SAMA was Phase II SAMA implemented to was implemented to install install improved passive ECCS suction strainers. Phase II SAMAs 042, improved passive ECCS suction strainers. Phase II SAMAs 042, 044, and 044, 045, which and 045, addition of recommend addition which recommend independent injection of independent injection systems to systems mitigate this to mitigate failure event, this failure were evaluated.

event, were evaluated.

LCS-STR-PG-8002B LCS-STR-PG-8002B 1.20E-04 1.20E-04 1.014 1;014 strainer BS-ECCS strainer ECCS BS- This term This represents failure term represents of core failure of spray and core spray RHR suction and RHR suction 80028 plugged 8002B plugged strainer BS-8002B. A Phase I SAMA was implemented to strainer BS-8002B. A Phase I SAMA was implemented to install install improved passive ECCS suction strainers. Phase II SAMAs 042, improved passive ECCS suction strainers. Phase II SAMAs 042, 044, 044, and 045, which and 045, addition of recommend addition which recommend independent injection of independent injection systems to systems mitigate this to mitigate failure event, this failure were evaluated.

event, were evaluated.

E.1-15 E.1-15

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

( SAMA Analysis 50-293-LR, 06-848-02-LR pngrim Nuclear Power Station Pilgrim Environmental Report Applicant's Environmental Operating License license Renewal Stage Table E.1-3 E.1-3 Correlation of Level 1 Risk Significant Terms Correlation Terms to Evaluated SAMAs .

Event Name Event Name Probability RRW Event Description Event Disposition DIsposition ADS-XHE-FO-XISI AD5-XHE-FO-X151 7.40E-03 1.013 Operator Operator fails to This term represents operator failure to manually open the SRVs perform emergency emergency for depressurization during medium medium LOCA. Phase II SAMAs,SAMAs, depressurization depressurization procedures and installation of improvement of procedures including improvement of during medium medium instrumentation to enhance the likelihood of success of operator instrumentation operator LOCA LOCA action inin response response to accident conditions, have already beenbeen implemented. SAMAs were recommended implemented. No additional Phase II SAMAs recommended for this subject.

EDG-ENG-FR-EDGB EDG-ENG-FR-EDGB 6.10E-03 1.013 1.013 Emergency diesel This term represents random failure of EDG-B.

EDG-B, leading to an SBO SBO generator generator -B -B (EDG)

(EDG) event. Phase I SAMAs to improve 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 SAMAs 025, 026, 027, 028, 029, 030, implemented. Phase IIII 5AMAs implemented.

033, and 035, for enhancing enhanCing AC or DC system reliability or to cope with LOOP and SBO events, were evaluated.

AC8-CBR-CO-104 AC8-CBR-CO-104 9.50E-05 1.013 480V circuit breaker circuitbreaker This term represents random random failure of 480V circuit breaker 52-52-104 fails to 104, MCC B17 and its 104, leading to loss of power to 480V MCCB17 remain closed associated loads. A Phase II SAMSAMA A was implemented implemented to proceduralize operator action to manually close the circuit proceduralize breaker. Phase IIII 5AMAs improve 480V bus SAMAs 030 and 058 to improve availability were evaluated.

HCI-MAI-MA-HCITM HCI-MAI-MA-HCITM 1.62E-02 1.013 HPCI unavailable This term represents HPCI system unavailable due to maintenance due to maintenance SAMAs to improve availability and maintenance. Phase I SAMAs reliability of the HPCI system that have already been been implemented raiSing backpressure trip setpoints and implemented include raising and operation. Additional improvements proceduralizing intermittent operation.

proceduralizing were evaluated in Phase II SAMAs 041, 042, 043, 044, and SAMAs 040, 041, 045.

E.1-16 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 Environmental Report Report Operating Operating License Renewal Stage E.1-3 Table E.1-3 Correlation Correlation of Level 1I Risk Significant Terms to Evaluated SAMAs Event Name Probability Probability RRW RRW Event Description Event Description Disposition Disposition SSW-CCF-FR-3P208 SSW-CCF-FR-3P208 5.59E-06 1.012 1.012 Common cause Common cause This term This term represents represents common cause failure common cause of 3 service water failure of failure of of 33 SSW pumps toto continue continue to run Phase Phase II SAMAs SAMAs were implemented implemented to pumps to run improve improve service water system reliability by enhancing screen wash, adding redundant DC control power power for SSW pumps, and increasing seismic integrity integrity of the partition wall between the the SSW SSW pumps.

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

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

evaluated .

IE-T3C

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

RBC-CCF-CC-4MOVS RBC-CCF-CC-4MOVS 1.13E-05 1.13E-05 1.012 Common Common cause This term represents common cause failure of RBCCW RBCCW heat failure of RBCCW RBCCW exchanger A & B B side MOVs MOVs to open. A Phase I SAMA SAMA was heat exchanger A & & implemented to improve RBCCW implemented RBCCW system reliability by making B

B side MOVs MOVs(4) (4)to component cooling water trains separate.

separate. Phase II II SAMA SAMA 055 to open improve RBCCW improve RBCCW system reliability reliability by reducing reducing common dependencies was evaluated.

E.1-17 E.1-17

Exhibit No. NRC000001 r

NRC - Applicant's Environmental Report r Pilgrim LR Proceeding Ar7 SAMA Analysis Ak

=- I-I-50-293-LR, 06-848-02-LR Ar Power Station Pilgrim Nuclear Power Environmental Report Applicant's Environmental Operating License Renewal Stage Table E.1-3E.1-3 Correlation of Level 1I Risk Significant Terms to Evaluated SAMAs Event Name Event Name Probability Probability RRW RRW Event Description Event Description Disposition Disposition OSP-24 OSP-24 1.41 E-02 1.41E-02 1.011 1.011 Failure to recover Failure recover This term This term represents represents operator failure to operator failure to recover offsite power power offsite power within

~ffsite power within 24 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> hours during during aa LOOP LOOP event.

event. PhasePhase II SAMAs, SAMAs, including 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> hours improvement of improvement sea of SBO procedures procedures and and training to enhance the likelihood of likelihood of success success of of operator action in in response to accident conditions, have conditions, have already been implemented. No additional Phase II11 SAMAs SAMAs were recommended for this subject.

SSW-RCI-FE-3828X SSW-RCI-FE-3828X 3.OOE-04 3.00E-04 1.01 1.01 Pressure switch Pressure This term represents random This random failure failure of SSWSSW pressure switch PS-PS-3828X coil PS-3828X coil fails 3828X, resulting in in loss loss of SSW system loop A. Phase II SAMAs SAMAs to energize energize were implemented implemented to improve improve service water system reliability by by enhancing screen wash, adding redundant DC control power for SSW pumps, pumps, and and increasing seismic integrity integrity of the partition wall between the SSW pumps.

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

evaluated.

EDG-MAI-MA-EDGA EDG-MAI-MA-EDGA 6.41E-03 6.41E-03 1.01 1.01 EDG-A EDG-A out for for This term term represents represents EDG-A EDG-A out for maintenance, maintenance, leading to an an maintenance maintenanCe sea SBO event. Phase I SAMAs SAMAs to improve availability and reliability of the EDGs of EDGs byby creating a a cross-tie of EDGs EDGs fuel oil supply and installing aa backup SBO sea diesel generator generator have already been been implemented.

implemented. Phase II II SAMAs SAMAs 025, 026, 027, 028, 029, 030, 026,027,028,029,030, 033, and 035, for enhancing AC or DC system reliability or to 033, cope with LOOP and SBO sea events, were evaluated.

E.1-18 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 Environmental Report Operating License Renewal Stage Stage Table E.1-3 E.1-3 Correlation of Level 1 Correlation I Risk Significant Terms Terms to Evaluated SAMAs SAMAs Name Event Name ' Probability RRW Event Description Disposition EDG-ENG-FR-EDGA EDG-ENG-FR';EDGA 6.1 OE-03 6.10E-03 1.01 EDG-A fails to This term represents This EDG-A, leading to an SBO represents random failure of EDG-A, continue to run SAMAs to improve availability and reliability of the event. Phase I SAMAs EDGs by creating aa cross-tie of EDGs fuel oil supply and installing a backup SBO diesel generator generator have already been implemented. Phase IIII SAMAs 025, 026, 027, 028, 029, 030, implemented.

033, and 035, for enhancing AC or DC system reliability or to cope with LOOP and SBO events, were evaluated.

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

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

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

SSW-MDP-FS-P208A SSW-MDP-FS-P208A 2.02E-03 2.02E-03 1.009 SSW pump P-208A This term represents random failure of SSW pump P-208A to fails to start on SAMAs were implemented start. Phase II SAMAs 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 between the SSW pumps.

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

E.1-19

Exhibit No. NRC000001 r

NRC - Applicant's Environmental Report Pilgrim LR Proceeding

- - de, SAMA Analysis

... ~-f::::-

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. I (Jr, 50-293-LR, 06-848-02-LR k-Pilgrim Nuclear Pilgrim Nuclear Power Station Station Environmental Report Applicant's Environmental Report Operating License Renewal Stage Stage Table E.1-3 Table E.1-3 Correlation of Level 1 I Risk Significant Terms Terms to Evaluated Evaluated SAMAs Name Event Name Probability RRW RRW Event Description Event Description Disposition Disposition C

C 5.80E-06 5.80E-OS 1.009 1.009 Reactor Protection Protection This term represel)ts represents failure failure of the RPS. Several Phase Phase I SAMAs SAMAs System (RPS)

(RPS) to minimize minimize the risks associated with anticipated transient transient without failure failure scram (ATWS) scenarios scram (ATWS) have already scenarios have been installed.

already been installed. No Phase No Phase II11 SAMAs were evaluated SAMAs evaluated to to further further improve improve reliability reliability of RPS.

RPS.

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

evaluated.

AC4-RCK-NO-605 AC4-RCK-NO-S05 2.51 E-03 2.51E-03 1.009 1.009 4.16kV circuit 4.1SkV circuit This term term represents represents failure of the control control circuit of 4.1SkV 4.16kV circuit breaker 152-605 152-S05 breaker 152-605, breaker 152-605, leading to to loss loss of power power toto safety busbus A6.

AS.

.. control circuit no no Phase I SAMAs SAMAs to improve 4.16kV 4.1SkV bus cross-tie capability and output procedures to procedures to repair or replace failed 4.1SkV 4.16kV breakers breakers have already been installed. In In addition, a a Phase I SAMA was implemented to proceduralize implemented proceduralize operator action to operator action to manually close the circuit breaker.

breaker. Phase Phase II II SAMAs SAMAs 025,025, 026, 02S, 027, 028,028, 029, 029, 030, 033, and 030, and 035 for enhancing enhancing AC AC or DC system reliability or to LOOP and seo cope with .LOOP SBO events were evaluated.

RCI-TDP-RS-P206 RCI- TDP-RS~P20S 1.52E-02 1.52E-02 1.009 1.009 RCIC turbine RCIC driven turbine driven This term This term represents random random failure of the RCIC system. Phase II pump P-20S pump P-206 fails fails to to SAMAs SAMAs to to improve improve availability and reliability reliability of of the RCIC RCIC system system restart after after clear that have have already already been implemented implemented include raising

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, . ~ -~

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high level high level signal signal backpressure trip backpressure trip setpoints setpoints andand proceduralizing intermittent proceduralizing intermittent

... operation. Additional improvements operation. Additional improvements were evaluated in were evaluated Phase II in Phase II

---

~

-,~

.. SAMAs SAMAs 040, 041, 042, 043, 044, 041, 042, 044, and and 045.

045 .

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

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

E.1-20 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 Station Applicant's Environmental Environmental Report Report Operating License Renewal Stage Table E.1-3 Table E.1-3 Correlation I Risk Significant Terms Correlation of Level 1 Terms to Evaluated Evah.lated SAMAs Event Name Name Probability RRW Event Event Description Disposition AC4-RCK-NO-508 AC4-RCK-NO-508 2.51 E-03 2.51E-03 1.008 4.16kV circuit This term represents failure of the control circuit of 4.16kV circuit breaker 152-508 152-508 152-508, leading to loss of power breaker 152-508, powerto to 480V load center B B1.

1.

control circuit no Phase I SAMAs SAMAs to improve 4.16kV bus cross-tie capability and output procedures to repair or replace failed 4.16kV breakers revise procedures have already been implemented.

implemented. In Inaddition, a Phase I SAMA was implemented implemented to proceduralize proceduraJize operator action to manually manually close the circuit breaker. Phase II II SAMAs 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 AC8-RCK-NO-101 2.50E-03 1.008 480V circuit breaker breaker This term represents random failure of 480V circuit breaker 52-52-101 control 101, leading to loss of power to 480V load center B BI1 and its circuit no ol:ltput output associated loads. A Phase I SAMA was implemented implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 030 and 058 to improve 480V bus II SAMAs availability were evaluated.

EDG-MAI-MA-EDGB EDG-MAI-MA-EDGB 4.09E-03 4.09E-03 1.008 1.008 EDG-B out for EDG-B EDG-B out for maintenance, leading to an This term represents EDG-B an maintenance SBO event. Phase I SAMAsSAMAs to improve improve availability and reliability of the EDGs by creating a cross-tie of EDGs fuel oil supply and a backup SBO diesel generator have already been installing a implemented. Phase IIII SAMAs 025, 026, 027, 028, 029, 030, implemented.

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 Applicant's Environmental Operating License Renewal Stage Table E.1-3E.1-3 Correlation Correlation of Level I1 Risk Significant Terms Terms to Evaluated SAMAs Event Name Event Name Probability RRW RRW* Event Description Disposition HCI-TDP-FS-PM205 HCI*TDP-FS-PM20S 7.53E-03 7.S3E-03 1.008 1.008 HPCI HPCI turbine driven This term represents represents random random failure of the HPCI HPCI system. Phase l system. Phase.1 pump P-20S P-205 fails to SAMAs to improve SAMAs improve availability and reliability of the HPCI system system start on demand demand that have already been implemented include raising backpressure trip backpressure trip setpoints setpoints and and proceduralizing proceduralizing intermittent intermittent operation. Additional improvements operation. improvements were evaluated in in Phase Phase"II SAMAs 040, 041, SAMAs 041, 042, 043, 044, and 04S. 045.

RBC-CCF-FS-4PUMP RBC-CCF-FS-4PUMP 7.35E-06 7.3SE-06 1.008 Common cause Common This term represents common cause failure of four RBCCW This RBCCW failure of four pumps to pumps to start. A A Phase Phase I SAMA SAMA was implemented implemented to improve RBCCW pumps RBCCW pumps to to RBCCW RBCCW system reliability reliability by by making component component cooling water water start trains separate.

trains separate. Phase" Phase II SAMA SAMA OS5055 to to improve improve RBCCW RBCCW system system reliability by reducing common reliability common dependencies dependencies was evaluated.

AC4-RCK-NO-505 AC4-RCK-NO-SOS 2.51 E-03 2.51E-03 1.007 1.007 4.16kV circuit This term represents failure of the control circuit of 4.16kV This 4.16kV circuit 152-505 breaker 152-5OS breaker 1S2-S0S, 152-505, leading to loss of power supply to safety bus control circuit cirCUit no A5.

AS. Phase II SAMAs 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. In Inaddition, aa Phase I SAMASAM A was implemented implemented to proceduralize operator operator action to manually manually close the circuit breaker.

breaker. Phase" Phase II SAMAs SAMAs 025, 026, 027, 028, 029, 025,026,027,028,029, 030, 033, and 035 03S for enhancing AC or DC DC system reliability or to cope with LOOP and SBO events were evaluated.

FXT-XVM-CC-511 FXT-XVM-CC-S11 5.OOE-04 S.00E-04 1.007 Manual valve 10 This term represents random failure of manual Thisterm manual valve 10-HO-511 HO-511 fails to HO-S11 fails to open open to provide fire water to LPClloops LPCI loops A and B. B. This failure failure open leads to loss of fire water backup for reactor vessel injection-injection and and drywell spray. Phase" Phase II SAMA SAMA 059OS9 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 Pilgrim Power Station Nuclear Power Station Applicant's Environmental Report Applicant's Environmental Report Operating License Operating Renewal Stage License Renewal Stage Table E.1-3 Table E.1-3 Correlation of Level 1 Risk Significant Terms Correlation of Level I Risk Significant Terms to Evaluated SAMAs to Evaluated SAMAs Event Name Event Name Probability Probability RRW RRW Event DescrIption Event Description Disposition Disposition FXT-XVM-CC-8156 FXT-XVM-CC-8156 5.00E-04 5.00E-04 1.007 1.007 Manual valve 8-1-56 Manual valve 8-1-56 This represents random term represents This term failure of random failure manual valve of manual 8-1-56 to valve 8-1-56 to fails to fails to open open open to open provide fire to provide water to fire water loops AA and LPCI loops to LPCI and B. This failure B. This failure leads to leads to loss loss of fire water of fire backup for water backup for reactor injection and vessel injection reactor vessel and spray. Phase drywell spray.

drywell Phase II SAMA 059 II SAMA 059 to to enhance availability of enhance availability of the the fire water fire system was water system was evaluated.

evaluated.

RCI-MAI-MA-RCITM RCI-MAI-MA-RCITM 11.97E-02

.97E-02 1.007 1.007 RCIC unavailable RCIC unavailable This represents RCIC term represents This term RCIC system unavailable due system unavailable due toto due due to maintenance. Phase maintenance maintenance.

to maintenance SAMAs to Phase II SAMAs to improve availability and improve availability and reliability of reliability the RCIC of the system that RCIC system that have already been have already been implemented include implemented include raising backpressure trip raising backpressure setpoints and trip setpoints and proceduralizing intermittent proceduralizing Additional improvements operation. Additional intermittent operation. improvements were evaluated in were evaluated Phase II in Phase SAMAs 040, II SAMAs 041, 042, 040, 041, 044, and 043, 044, 042, 043, and 045.

045.

CIV-AOV-CC-5042B CIV-AOV,:CC-5042B 1.OOE-03 1.00E-03 1.007 1.007 AO fails to 5042B fails AO 5042B to This represents random term represents This term failure of random failure of DTV valve AO DTV valve 5042B to AO 5042B to open on demand open on demand open on demand, resulting in loss of open on demand, resulting in loss of containment ventingcontainment venting capability to control capability pressure. Phase containment pressure.

control containment Phase II SAMAs 001, II SAMAs 001, 009, 014, and 059, to provide alternate 014, and 059, provide alternate means of suppressionmeans of suppression cooling and drywell pool cooling pool spray and drywell spray and to enhance the to enhance availability and the availability and reliability of firewater for reactor vessel injection and drywell reliability of firewater for reactor vessel injection and drywell spray, evaluated for were evaluated spray, were containment pressure for containment pressure control.

control.

CIV-AOV-CC-A5025 CIV-AOV-CC-A5025 1.OOE-03 1.00E-03 1.007 1.007 AO 5025 fails AO 5025 fails to to This represents random term represents This term failure of random failure of DTV valve AO DTV valve 5025 to AO 5025 to ..

open on open demand demand open on demand, resulting in loss of open on demand, resulting in loss of containment ventingcontainment venting capability to to control pressure. Phase containment pressure.

control containment Phase II SAMAs 001, II SAMAs 001, 009, 014, 009,014, and 059, to provide alternate means 059, provide alternate means of suppression of suppression pool cooling and and drywell spray and drywell spray and to enhance the to enhance availability and the availability and reliability of firewater for reactor vessel injection and drywell reliability of firewater for reactor vessel injection and drywell spray, were evaluated for were evaluated containment pressure for containment pressure control.

control.

E.1-23 E.1-23

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

1. Analysis le",

-. -7.--.- 7--1-7-1a-I NK c

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Applicant's Environmental Report Renewal Stage Operating License Renewal Table E.1-3 E.1-3 of Level 1I Risk Significant Terms Correlation of Evaluated SAM Terms to Evaluated SAMAs As Name Event Name Probability RRW Event Description Disposition CM 3.30E-01 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 SAMAs already been installed. No Phase IIII SAMAs were evaluated to further improve reliability of RPS. However, Phase IIII SAMA SAMA 048 to enhance reliability of the standby liquid control system and and improve ATWS capability to mitigate the consequences of this event was evaluated.

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

RBC-MAI-MA-P202F RBC-MAI-MA-P202F 6.44E-03 6.44E ...03 . 1.006 RBCCW RB.CCWpump pump This term represents RBCCW pump 202F unavailable represents RBCCW unavailable due to 202F 202Foutout for maintenance.

maintenance. A Phase I SAMA SAMA wasWas implemented implemented to improve maintenance maintenance RBCCW system reliability by making component cooling water RBCCW water trains separate.

separate. Phase IIII SAMA SAMA 055 to improve RBCCW system improve RBCCW reliability by reducing reducing common common dependencies dependencies was evaluated.

evaluated.

IE-TDC-CCF IE-TOC-CCF 3.66E-08 3.66E-08 1.006 Common Common cause This term represents represents an initiating event caused by a complete failure of 125VOC125VDC loss of 125VOC 125VDC buses D-160-16 and D-17 0-17 or random random failure of of buses A&BA&B batteries 0-1 D-1 and and D-2.

0:"2. Phase Phase I SAMAs SAMAs to improve improve battery battery charging capability and replace existing batteries with more reliable ones have already already been installed. Phase II II SAMAs SAMAs 025, 026, 027, 031, 032, 033, 034, and 035 for enhancing DC system 026,027,031,032,033,034, system availability and reliability were evaluated.

E.1 -24 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 Environmental Report Operating License Renewal Stage E.1-3 Table E.1-3 Correlation I Risk Significant Terms to Evaluated SAMAs Correlation of Level 1 Event Name Probability RRW Event Description Disposition SPC-MAI-MA-SPCA SPC-MAI-MA-SPCA 3.01 E-03 3.01E-03 1.005 Suppression pool This term represents RHR suppression pool cooling loop A cooling loop A out unavailable due to maintenance. SAMAs to improve maintenance. Phase I SAMAs for maintenance availability and reliability of the RHR suppression pool cooling mode that have already been implemented mode implemented include using drywell drywell spray mode and fire protection cross-tie to provide redundant containment heat removal containment removal capability. Additional improvements were evaluated in in Phase IIII SAMAs SAMAs 001 and 014.

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

DWS-MAI-MA-DWSA OWS-MAJ-MA-OWSA 3.18E-03 3.18E-03 1.005 Drywell spray loop This term represents RHR drywell spray loop A unavailable due A out for maintenance. Phase I SAMAs to maintenance. 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 implemented fire protection cross-tie to provide redundant containment heat heat improvements were evaluated in removal capability. Additional improvements Phase IIII SAMA 009.

E.1-25 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 Applicant's Environmental Environmental Report Report Operating License Renewal Stage Stage Table E.1-3 Table E.1-3 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs Event Name Event Name Probability RRW Event Description Disposition ADS-XHE-FO-XIS2 ADS-XHE-FO-X1 S2 1.45E-03 1.45E-03 1.005 Operator fails to This term represents operator failure to manually open the SRVs perform emergency for depressurization during aa small LOCA. Phase I SAMAs, depressurization including improvement of procedures and installation of of during small LOCA LOCA instrumentation to enhance the likelihood of success of operator action in in response to accident conditions, have already been implemented.

implemented. No additional Phase" Phase IISAMAs SAMAs were recommended 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 Applicant's Environmental Report Operating License Renewal Stage L.)

E.1.2 PSA Model - Level 2 Analysis Containment Performance E.1.2.1 Containment Performance Analysis The PNPS Level 2 PSA model used for the SAMA 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.t-3]. E.1-3].

The Level 2 PSA model used for the SAMA SAMA analysis, Revision 1, 1, reflects the PNPS operating configuration and design changes as of September September 2001.2001. Specifically, the Level 2 model has been updated to incorporate incorporate insights from the independent independent BWROGBWROG peer review.

The PNPS Level 2 model includes two types of considerations: (1) (1) a deterministic deterministic analysis of the physical processes for a spectrum of severe accident progressions, 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 performed by

• • utilization of the MAAP MAAP code [Reference

[Reference E. 14] to simulate severe accidents that have E.1-4] 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 Examples include debris coolability, pressure spikes due to ex-vessel steam explosions, scoping calculation of direct containment heating, molten molten debris filling the pedestal sump and flowing over the drywell floor, ( )

containment bypass, deflagration and detonation of hydrogen, thrust forces at reactor reactor '-'

vessel failure, liner melt-through, and thermal attack of containment penetrations.

The Level 2 analysis examined examined the dominant accident sequences and the resulting plant damage damage states (PDS)

(PDS) defined in Level 1. 1. The Level 1 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. 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) (CET) with functional nodes that represent represent phenomenological events*andevents 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 Frequency (LERF) performance from the containment performance Level 2 results because the magnitude and timing of these releases provide the greatest 2'results greatest nntontfil fnr nnt"nti~1 incriv fnr ,,~rI\I haIth "ff"rt~

h"!:Ilth affontc tn theg nohlh-tn th" Tha frong n"hlir Th" icnev role fronl,,,n,,,, infin*nnrnyimqtcmlv r!:llr"I!:It"d i~ QnnrnYim~t"I\l

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Nuclear Power Station Applicant's Environmental Environmental Report kmr' License Renewal Stage Operating license Table E.1-4 E.1-4 Summary of PNPS PSA Core Damage Summary Damage Accident Class PoS PDS S l Point

'Point,  % of Total

% of Total Group Simplified Description SipiidDescription Estimate CDF Group Estimate CDF LOCAs LOCAs LOCA with initial or long-term loss Large and small break LOCA 1.16E-7 1.16E-7 1.80 1.BO of core cooling. Core damage results at low or or high high For most POS, reactor pressure. For PDS, late injection and late injection and containment containment heat removal are available.

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

seo SBO seo SBO involving aa loss of high-pressure high-pressure injection. Core 1.48E-7 1.4BE-7 2.31 (stuck-open SRV) or high damage results at either low (stuck-open reactor pressure. All accident mitigating functions are power is restored.

recoverable when AC power VSLRUPT VSL_RUPT Vessel rupture event resulting In In LOCA beyond ECCS ECCS 4.OOE-9 4.00E-9 0.06 0.06 capability. All All PDS in core result in POSresult core damage damage at at low low reactor reactor pressure with late injection available.

ATWS ATWS Short-term Short-term ATWS that leads to early core damage at high 3.39E-8 3.39E-B 0.53 0.53 reactor pressure following loss of reactivity control and rapid containment pressurization.

pressurization. Reactor coolant system leakage rates associated with boil-off of coolant through the cycling of SRVs/SV SRVs/SV with early core melt subsequent to containment overpressure failure: failure. Late Late injection and containment containment heat removal are available.

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

TW Containment decay heat removal systems are not Containment 5.86E-6 '

5.B6E-6 91.45 available and coolant recirculation to the torus over pressurizes the containment to failure or venting. The  ;

is saturated.

torus is

Total Total 6.41 E-06

,6.41E-06 1.OOE+00 1.00E+OO 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 Environmental Report {,\

Operating License Renewal Stage Stage (

' - 'I Table Table E.1-5E~1-5 Notation and Definitions for PNPS CET Notation CET Functional Nodes Description CET CET NodeNode CET Functional Node Description CET Description Plant Damage Plant Damage State This top event represents the initiators considered in the containment containment (PDSEVNT)

Event (PDS_EVNT) performance performance analysis.

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

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

(RPV@VF)

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

Cooling In-Vessel Cooling In-Vessel This top event addresses the recovery of coolant injection into the vessel (IN-REC)

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

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

Early Containment Early Containment This top event node node considers considers thethe potential potential loss loss of containment containment integrity integrity at, Failure (CFE)

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 phenomena are containment isolation failure; containment bypass; containment overpressure failure at vessel breach; hydrogen deflagration or detonation; detonation; fuel-coolant interactions (steam(steam explosions); high pressure melt ejection and subsequent direct containment heating; and drywell heating; drywell steel steel shell melt-through.

melt-through.

Early Release to This top event node considers the importance importance of early torus pool scrubbing Torus (EPOOl)

(EPOOL) in mitigating the magnitude magnitude of fission products released from the damaged damaged core. Success implies that fission product transport path subsequent to early containment containment failure is through the torus water and the torus airspace.

That is, the torus pool Is is not bypassed.

bypassed. Failure involves a release into the drywell.

Debris Debris Cooled Ex- This top event considers the delivery of water to the drywell, via drywell drywell vessel (DCOOl)

(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 molten core attacks concrete in the reactor pedestal, that core debris remains hot, and sparing sparing of the concrete decomposition products through the melt releases the less volatile fission products to the containment atmosphere 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 Power Station Applicant's Environmental Environmental Report Operating License license Renewal Stage Table E.1-5

• Table E.1-5 Definitions for PNPS CET Notation and Definitions CET Functional Nodes Description (Continued)

(Continued)

CET CET Node CET Functional Node Description Description Late Containment This top event addresses the potential loss of containment integrity in the Failure .(CFL)

(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 non-condensable concrete.

concrete; Late Release to This top event node considers the importance importance of late torus pool scrubbing in in Torus (LPOOL) mitigating the magnitude of fission products released from the damaged 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 Failure involves a release into the drywell.

Fission Product This top event addresses fission product releases from the fuel into the (FPR)

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

Reactor Building This top event is used to assess the ability of the reactor building to retain (RB)

(RB) fission products released from containment. Success of top event RB is is defined to be a reduction of the containment release magnitude.

magnitude.

E.1-30*

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

'-')'

l.\

Ealy Medun Ea1y Release MldillTl Rlease Eary LowFRlease , mo 11.01%

1°/

.0.52% a '/uly HAh Release Ealy Hg, Release LateeHgh Hg, Reease Release .76%

1.7SOA.

Lat, 0.00%

O4.

OO/o a\

Eal/ -No Con~imnat Failure Late MltillTl MWc Release .

Jum Release- 1.73%

24.33"10 2A AN Late Release LateLI..o.v R.lease 70.65%

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

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

Transierts Interfaang Interfacing System System LOCAs LOCAs LOCus Aticipated 2 LOCAs Anticipated Transiert Transient wthot

'AithouI 201%

2.01 % 13 YO t1.13% 0.01%

Scram Sacam 0.01%

39.82%

39.82% Vesa Rome 0.01%

1. .

Statioo Staficn Blackout 57.03%

57.03%

Figure Figure E.1-2 E.1-2 PNPS Plant Damage Damage State Contribution 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 Applicant's Environmental Operating license License Renewal Stage L)

E.1.2.2 Radionuclide Analysis E.1.2.2.1 Introduction A major feature of a Level 2 analysis is the estimation estimation of the source term for every possible outcome outcome of the CET. The CET end points represent the outcomes of possible in-containment in-containment accident progression sequences. These end points represent complete severe accident sequences from initiating event to release of radionuclides to the environment. environment. The Level 1 I and information is passed through plant system information 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 number of postulated accident scenarios considered, mechanistic calculations (i.e., (Le., MAAP calculations) are not performed performed for every end-state ;n in the CET.

CET. Rather, accident sequences produced by the CET CET are grouped or "binned"

'binned' into a limited numbernumber of release categories each of of which represents all postulated accident scenarios that would produce a .similar similar fission product source term.

The criteria used to characterize the release are the estimated magnitude 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 determined using uSIng the results of MAAPMAAP calculations [Reference

[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 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).

hours).

Based on MAAP MAAP calculations for a spectrum of severe accident sequences, PNPS expects that Emergency Action Level (as defined by the PNPS Emergency an Emergency Emergency Plan)

Plan) will be reached within the first half hour after accident initiation. Reaching an EmergencyEmergency Action level Level initiates a formal 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 deCision-making 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 CETCET 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 /> measuredmeasured from the time of of accident initiation and for which minimal minimal offsite protective measures would be accomplished.

• • Late releases are CET CET end-states involving containment failure greater 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 measures are fully effective.

E.1-33 u

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Pilgrim Nuclear Power Station Station Applicant's Environmental Report Applicant's Environmental Report License Renewal Stage Operating License Stage E.1.2.2.3 E.1.2.2.3 Magnitude of Release Magnitude ,

Source term results from previous categorization of release magnitude previous risk studies suggest that categorization magnitude based on cesium iodide (CsI) appropriate [Reference E.1-5].

(CsJ) release fractions alone are appropriate E.1-5]. The CsI Csi release fraction indicates the fraction of radionuclides escaping to the environment.

o( in-vessel radionuclides environment.

releaselevels are (Noble gas release'levels non-informative since release of the total core inventory are non-informative inventory of noble gases is essentially complete given containmentcontainment failure).

The source terms were grouped into four distinct radionuclide release categories or bins release categories according to release magnitude according magnitude as follows:

(1) (HI) - A radionuclide release of sufficient magnitude High (HI) magnitude to have have the potential to cause early fatalities. This implies a total integrated release of >10% >10% of the initial core inventory of Csi E.1-5].

[Reference E.1-5].1 1 Csl [Reference (2)

(2) Medium (MED)

Medium (MEO) - A radionuclide release of sufficient magnitude magnitude to cause near- near-term health effects. This implies a total integrated release of between between 1 and and 10% of the initial core inventory of Csi 10% CsI [Reference E. 1-5].2

[Reference E.1-5].2 (3)

(3) (LO) - A radionuclide release with the potential for latent health effects. This Low (LO)-

implies a total integrated release of between 0.001% 0.001 % andan'd 11%

% of the initial core inventory of Csl. CsI. ' '

(4)

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

<0.001%

<0.001  % of the initial core inventory of Csl.

release" as used in the above categories is defined as the integrated integrated release" The "total 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 "total integrated 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.

release" E.1.2.2.4 ' Release Category Bin Assignments .

summarizes the scheme used to bin sequences with respect to magnitude Table E.1-6 summarizes magnitude of of release, based on the predicted CslreleaseCsl release fraction and release timing. The combi~atiqn combi nation of of '

magnitude and timing produce seven distinct release categories for source terms.

release magnitude terms. These are the representative release categories presented in Table E.1-7. E. 1-7. I ' ,

1. Once the Csi

1. Csl sou~ce source term exceeds 0.1" 0.1, the source term is Islarge large enough 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 for'Csl number of Csi release fractions of 11 to 10%, the number be at least 10% of the latent fatalities for the highest release.

is found to be latent fatalities is 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 Nuclear Power Station Applicant's Environmental Applicant's Report Environmental Report {"I Operating Operating License Renewal Stage ~

Table E.1-6 TableE.1-6 Release Severity and Timing Classification Scheme Release Scheme Summary Summary

, Release Severity Release Timing Classification Classification Time of Initial Release from Csi % Release Category Csl %Release Category Category Accident Initiation High Greater than 10 10 (E)

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 Medium i1 to to 10 Low 0.001 to 11 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 Table E.1-7

~NPS Release PNPS Release Categories Categories Timing of Timing of Magnitude Magnitude of Release of Release Release Low Low Medium Medium High High Early Late Early/Low Early/Low Late/Low Late/Low Early/Med Early/Med Late/Med Early/High Early/High Late/High NCF NCF Cu )

E.1.2.2.5 E.1.2.2.5 Mapping Mapping of Level 1 Results into the Various Release Categories Categories PDS provide the interface between the Level 11 and Level 2 analyses (i.e. between core damage 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 important to core damage in the Level 1 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 initial conditions that are likely to yield a similar accident progression through the Level 2 CETs and the attendant attendant challenges to containment integrity.

From From the perspective Perspective of the Level 2 assessment, assessment, PDS binning entails the transfer of specific information information from the Level 1 to the Level 2 analyses.

• Equipment failures in Level 1. 1. Equipment Equipment failures in support systems, accident prevention systems, and mitigation systems that have been noted 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 consid~ration consideration of E.1-35 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 Environmental Report Operating Operating License Renewal Stage appropriate, has been performed as part of the Level 2 environments where appropriate, adverse environments analysis. .

• • RPV status. The RPV pressure condition is explicitly transferred from the Level 1 I analysis to 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.

damage.

• • Accident sequence timing. Differences in accident sequence timing are transferred transferred with the Level 1 1 sequences. Timing affects such sequences as SBO, internal flooding, and containment bypass (ISLOCA).

(ISLOCA). .

This transfer of information allows timing to be properly assessed in the Level 2 analysis.

Based on the above criteria, the Level 11 results were binned into 48 PDS. These PDS define important combinations of system states that can result in distinctly differentdifferent 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 E.1-8 Table E.1-8 Summary of PNPS Summary PNPS Core Damage Accident Sequences Core Damage Sequences Plant Damage Damage States PDS Description Description Point Estimate %fD

%ofCDF

%ofCDF Estimate PDS-1 Long-term LOeA Long-term high-pressure core makeup LOCA with loss of high-pressure O.OOE+00 O.OOE+OO 0.00 0.00 from HPCI and RCIC, loss of containment~eat 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 low-pressure firewater) is systems (core spray, LPCI, and firewater) is available, primary system depressurization occurs. The provided primary containment isis vented and intact.

PDS-2 Long-term Long-term LOCA with loss of both high-pressure core 1.05E-11 1.0SE-11 <0.001

<0.001 makeup (HPCI (HPCI and RCIC) and containment heat removal.

Core damage results at high primary system pressure.

containment venting fails, containment failure Because containment failure ,

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

E.1-36 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 Power Station Report Applicant's Environmental Report Operating License Renewal Stage Operating U

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

(Continued)

Point PDS Description Estimate Estimate

%of CDF

%ofCDF PDS-3 Short-term LOCA with loss of high-pressure core makeup' makeup, 8.68E-08 S.GSE-OS 1.35 1.35 and failure to depressurize the primary system for low-pressure core makeup.

makeup. Core damage damage occurs at high primary system pressure. Late injection from core spray, primary LPCI, LPCI, and firewater is is available, provided primary system system

, depressurization occurs. Containment heat removal is is available.

PDS-4 Short-term LOCA with loss of high-pressure Short-term high-pressure core makeup, makeup, O.OOE+00 O.OOE+OO <0.001 loss of containment containment heat removal, and failure to depressurize primary system for low-pressure core depressurize the primary

. makeup.

makeup. Core damage occurs at high primary system system pressure. Late injection from core spray, LPCI, and firewater is available, provided primary system system depressurization occurs. Unlike PDS-3, depressurization PDS-3, containment heat containment heat removal isis unavailable.

PDS-5 Long-term LOCA with loss of high-pressure core makeup Long-term and containment containment heat removal. Core damage occurs at low primary system. Late injection is available from at low-trol)1low-0.OOE+00 O.OOE+OO 0.00 uQ.

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

PDS-6 PDS-G Long-term large LOCA. High-pressure core makeup from Long-term 0.00E+00 O.OOE+OO 0.00 0.00*

HPCI and RCIC HPCI RCIG are unavailable due to the large LOCA. LOCA.

Because containment venting fails, containment failure occurs long-term.

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

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

PDS0-PDS-8 Short-term large Short-term large LOCA LOCA with loss of core cooling. Core 4.43E-09 4.43E-09 0.07 damage results at low primary system pressure. Late injection from firewater cross tie is is available. However, However, PDS-7, containment heat removal is unlike PDS-7, is unavailable.

unavailable.*

E.1-37 u

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

(Continued)

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

• 0.06%

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

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

pressure.

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

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

PDS-13 Transient with a loss of long-term long-term decay heat removal. Core 3.75E-06 3.75E-06 58.5 damage results at high primary system pressure.

pressure. Late in-vessel and ex-vessel injection is available. Unlike PDS-12 PDS-12 containment venting fails.

PDS-14 Short-term transient with failure to depressurize the primaryprimary 1.52E-07 1.S2E-07 2.37 damage results at high primary system system. Core damage system pressure.

pressure. Late in-vessel and ex-vessel injection is is available. Containment heat removal from RHR is is  !

available. ,

PDS PDS-1S Short-term transient with failure to depressurize the primary Short-term primary 5.07E-08 S.07E-08 0.79 system. Core damage results at high primary system i

pressure. Late in-vessel and ex-vessel injection is is i Containment heat removal from RHR is available. Containment available. However, containment venting is not available. I 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 Environmental Report Operating License license Renewal Stage U I E.1-8 Table E.1-8 Summary of PNPS Core DamageDamage Accident Sequences Plant Damage Damage States (Continued)

(Continued)

Point PDS Description Point  % of CDF

%ofCDF Estimate PDS-16 Short-term transient with failure to depressurize the primary Short-term 4.89E-09 0.08 system. Core damage damage results at high primary primary system system pressure. Late in-vessel and ex-vessel injection is 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 Short-term primary 2.53E-09 0.04 damage results at high primary system system. Core damage pressure. Late in-vessel and ex-vessel injection is 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 1.56E-06 24.40 Core Core damage results at low primary system pressure. Late in-vessel and ex-vessel injection is available.

available .. The containment is vented and remains remains intact at the time of core damage.

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

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

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

injection and containment heat removal are available.

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

However, containment venting is not available.

PDS-231 PDS-23 Short-term Short-term transients with loss of core cooling. Core O.OOE+00 O.OOE+OO 0.00 damage results at low primary system pressure. Late injection and containment venting are available, but containment containment heat removal is not available.

PDS-24 Similar to PDS-23, PDS-23, except that containment venting is not 4.98E-09 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 Nuclear Power Pilgrim Nuclear Power Station Applicant's Environmental Report Applicant's Environmental Report Operating License Renewal Stage Operating Stage Table Table E.1-8 E.1-8 PNPS Core Damage Summary of PNPSCore Summary Damage Accident Sequences Plant Damage Accident Sequences Damage States (Continued)

(Continued)

Point Point PDS Description Description , Estimate Estimate C of CDF

%ofCDF PDS-25 Short-term transients with loss of core Short-term core cooling. CoreCore 2.57E-09 2.57E-09 0.04 primary system pressure. No late damage results at low primary injection, but containment heat removal removal and containment containment venting are available.

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

PDS-27 Short-term transients with loss of core cooling. Core Short-term 4.40E-11 0.001 damage results at low primary system pressure.

damage pressure. Late injection and containment containment heat removal are not available.

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

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

restored PDS-30 PDS-30 Short-term seo Short-term SBO sequence involving aa loss of of high-pressure high-pressure O.OOE+00 O.OOE+OO 0.00 injection at high primary system pressure from loss of all AC power power and DC DC power or or failure failure of of SRVs. All All accident-accident- "

mitigating functions are recoverable when offsite power power is is restored.

PDS-31 Long-term Long-term SBO seo sequence involving Involving aa loss of1high-pressure of high-pressure 2.60E-09 0.04 injection due to one stuck-open safety relief valve or long-term failure term of HPCI failure of and RCIC HPCI and RCIC andand subsequent failure to sub~equEmt failure to depressurize the primary system. Core damage results at at accident-rnitigating low primary system pressure. All accident~rnitigating functions are recoverable when offsite powe~ power is isrestored.

PDS-32 PDS-32 Short-term seo Short-term SBO sequence involving aa loss of,high-pressure of high-pressure 4.00E-09 4.OOE-09 0.06 injection due to two stuck-open safety relief valves or failure '

of HPCI and RCIC RCIC and one stuck-open safety relief valve.

Core damage results at primary system pressure. All at low primary recoverable when offsite accident-mitigating functions are recoverable offsite power is 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 Environmental Report license Renewal Stage Operating License U Table E.1-8 E.18 Summary Summary of PNPS Core Core Damage Accident Sequences Plant Damage Damage States (Continued)

(Continued)

PDS Description Description Point Estimate  % of CDF

%ofCDF Estimate PDS-33 Short-term large reactor vessel rupture. The resulting loss Short-term 4.OOE-09 4.00E-09 0.06 of coolant is beyond the makeup makeup capability of ECCS. Core damage occurs in the short term at low primary system pressure. Vessel injection and all forms of containment containment heat removal (RHR and containment venting) are available.

The containment is not bypassed and AC power is is available.

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

PDS-35 Short-term large reactor vessel rupture. The resulting loss Short-term O.OOE+00 O.OOE+OO 0.00 of coolant is beyond the makeup capability of ECCS. 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 containment is not bypassed PDS-36 PDS-36 and AC power is available.

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

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

PDS-38 Short-term Short-term ATWS ATWS that leads to early core damage at low 0.OOE+00 O.OOE+OO 0.00 primary 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 ATWS that leads to late core damagedamage at low 0.OOE+00 O.OOE+OO 0.00 primary primary system pressure following successful reactivity control. Late injection is available; containment containment heat removal removal from the RHR is not available. The containment containment isis vented.

E.1-41 E.1-41 C

'-,/J

{ "

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

(Continued)

Point PDS Description Description Estinate  % of CDF

%ofCDF*

Estimate PDS-41 PDS-41 Short-term ATWS that leads to early core damage Short-term damage at high 1.34E-11 1.34E-11 <0.001

<0.001 primary primary system pressure following successful reactivity control. Late injection and containment containment heat removal are available.

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

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

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

PDS-45 Short-term Short-term ATWS ATWS that leads to containment failure and 3.39E-08 3.39E-08 0.53 early core damage 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 PDS-46 Short-term Short-term ATWS ATWS that leads to containment containment failure and 0.OOE+00 O.OOE+OO 0.00 early core damage at high primary system pressure because of inadequate reactor water level following successful reactivity control. No late injection; however, successful.reactivity containment venting Is Is available.

PDS-47 Unisolated Unisolated LOCA LOCA outside containment with early core melt 3.22E-09 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 U \

Operating License Renewal Stage The PDS designators listed in Table E.1-8 represent the core damage end state categories from the Level 11 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 CET to determine the appropriate release category for each Level 2 sequence. 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, however, that since not all the Level 2 sequences associated with each Level Level 11 core damage class may be assigned to the same release category, there is no direct link between a specific Level 1 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 overall frequency of that release category. The CET CET described in the Level 2 model determines the release category frequency attributed to each Level 11 core damage PDS.

E.1.2.2.6 Collapsed Accident Progression E.1.2.2.6, Progression BinsSins Source Terms The source term analysis results in hundreds of source source terms for internal initiators, making calculation with the MACCS2 MACCS2 consequence model model cumbersome.

cumbersome. Therefore, Therefore, the source terms were grouped into a much smaller number number of source term groups defined in terms ,of of similar properties, with a frequency weighted mean source term for each group.

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

0 interactions. Descriptions of the CAPS CAPB are presented in TableE,1-9.

Table E.1-9.

E.1-43 E.1-43 U

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 Applicant's Environmental Report Report Operating license License Renewal Stage Table E.1-9 E~1-9 Collapsed Accident Progression Bins (CAPB) (CAPB) Descriptions CAPB ,

CAPS Description Number Number Description CAPB-1 CAPB-1 [CD, No VB,VB, No CF, CF, No CCI]

CCI]

Core damage (CD) (CD) occurs, but timely recovery of RPV injection prevents vessel vessel breach (No VB).

VB). Therefore, Therefore, containment integrity is not challenged (No CF) 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 CAP B-2 [CD, VB, No CF, No CCI] CCI]

Core damage (CD)

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

phenomena [DCH]) occurs, containment does not fail.

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

damage (CD)

Core damage (CO) occurs followed by vessel breach (VB). (VB). Containment does not fail structurally vented (No CF).

structurally and is not v~nted CF). ' However, However, ex-vessel releases releas~s are not recovered in time, and therefore core-concrete interactions occur (CCI). RPV RPV pressure is not important important because, even though high pressure induced severe phenomena (such accident phenomena (such as direct containment [DCH]) occurs, containment heating [OCH))

containment does not fail, nor is the vent limit reached.

CAPB-4 CAPB-4 [CD, VB, Early CF, WW, RPV pressure >200 psig at VB,

[CD, \/B, No CCII CCI)

Core damage (CD) (CD) occurs followed by Vessel

'vessel breach (VB). Containment fails either*,

(VB). Containment either II before core damage, damage, during core damage, damage, or at vessel breach (Early(Early CF).

CF). i i Containment failure occurs in the torus (WW), above the water level. level., RPV pressure Is greater greater than 200200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena phenomena [OCH][DCH] are possible). There are no core concrete interactions (No CCI) concrete' CCI) due to the' presence of an overlying pool of water.

E.1-44 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 Power Station Applicant's Environmental Report (,

Operating License Renewal Stage C!

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

(Continued)

CAPB CAPB Description Description Number Number CAPB-5 CAPB-5 [CD, VB, Early CF,

[CD. VB. CF. WW, WW. RPV pressure <200 psig at VB. VB, No CCI]

CCI]

Core damage damage (CD)

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

Containment failure occurs in the torus (WW). (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) CCI) due to the presence of an overlying pool of water.

CAPB-6 CAPB-6 [CD, VB,

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

VB, CCI]

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

(Early CF).

Containment Containment failure occurs in the torus (WW), above the water level. RPV pressure is greater greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena lOCH] [DCH] are possible). Following containment CAPB-7 failure, core-concrete interactions occur (CCI).

failure.

[CD,

[CD, VB, Early CF, WW.

(CCI).

WW, RPV pressure <200 psig at VB, CCI]

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

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

phenomena. Following Following containment failure, core-concrete interactions occur (CCI).

(CCI).

CAPB-8 CAPB-8 DW, RPV pressure >200

[CD, VB, Early CF, OW, >200 psig at VB, No CCI]

CCI]

Core damage damage (CD)

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

Containment Containment failure occurs in the drywell or below the torus water line (DW). (OW). RPV RPV greater than 200 psig at time of vessel breach.(this pressure is greater breach (this implies that high

, pressure induced severe accident phenomena [DCH] are possible). There are no phenomena lOCH] no core concrete interactions (No CCI) CCI) due to the presence of an overlying pool of of water.

E.1-45 E.1-45 u

---

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

(Continued)

(Continued)

CAPB Description Description Number Number CAPB-9 CAPB-9 VB, Early CF, DW,

[CD, VB, OW, RPV pressure <200 psig at VB, No No CCI]

CCI]

Core damage (CD)

Core (CD) occurs followed by by vessel breach breach (VB).

(VB). Containment Containment fails either damage, during core damage,or before core damage,during damage, or at vessel breach (Early CF). CF).

Containment failure occurs in Containment in the drywell or below the torus water line (OW). (DW). RPV RPV pressure is pressure isless than than 200 psig psig atat time time of of vessel vessel breach; breach; precluding precluding highhigh pressure pressure induced severe accident phenomena.

phenomena. There are no core concrete interactions (No (No CCI)

CCI) due to the presence of an overlying pool of water.

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

VB, CCI]

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

Containment Containment failure occurs in the drywell or below the torus water line (OW). (DW). RPV RPV pressure is greater than 200 psigpsigat at time of vessel breach (this implies that high pressure induced severe accident pressure accident phenomena phenomena [OCH1

[OCH] are are possible).

possible). Following containment failure, core-concrete interactions occur (CCI). (CCI).

CAPB-11 CAPB-11 [CD, VB, Early CF, OW,

[CD, DW, RPV pressure <200 psig at VB, VB, CCI]

CCI]

Core damage (CD) occurs followed by vessel breach (VB).

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

Containment Containment failure occurs in in the drywell or below the torus water line (J)W). (DW). RPV RPV pressure isis less than 200 psig at time of vessel breach; precluding high pressure Induced severe accident induced severe accident phenomena.

phenomena. Following Following containment containment failure, core-concrete interactions occur (CCI).

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

Core damage Core damage JCD)

{CO) occurs followed followed by by vessel vessel breach breach (VB).

(VB). Containment Containment fails fails late 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 high-pressure severe accident phenomena (such (such as as DCH)

OCH) did not not fail fail containment.

containment.

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

E.1-46 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 Power Station Applicant's Applicant's Environmental Environmental Report Operating License license Renewal Stage L)

(I Table E.1-9 Table *E.1-9 Collapsed Accident Progression Bins (CAP (CAPB) B) Descriptions (Continued)

(Continued)

CAPB Description Description Number Number CAPB-13 CAPB-13 [CD, VB, Late CF,

[CD. VB. WW, CCIX CF. WW. CCI]

Core damage (CD)(CD) occurs followed by vessel breach (VB). Containment Containment fails late (late CF)

CF) due to core-concrete interactions (CCI) Containment (CCI) after vessel breach. Containment failure occurs inin the torus (WW).

(WW), above the water level. RPV pressure is is not important because high-pressure severe accident phenomena important phenomena (such as DCH) OCH) did not fail containment. '

CAPB-14 CAPB-14 [CD, VB.

[CD. VB, Late Late CF, CF. DW, No CCI]

OW. No CCI]

Core damage (CD)(CD) occurs followed by vessel breach (VB). Containment fails late (VB). Containment due to loss of containment heat removal (Late CF). Containment Containment failure occurs in the drywell or below the torus water level (DW). (OW). RPV pressure is is not important because high-pressure because high-pressure severe accident accident phenomena did not not fail containment.

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

CAPB-1 CAPB-155 [CD, VB,

[CD. CF, OW.

VB. Late CF. DW, CCI]

CCI]

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

CAPB-16 CAPB-16 [CD, VB,

[CD. VB. BYPASS, BYPASS. RPV pressure >200 psig. psig, No CCI]

CCI]

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

CAPB-17 CAPB-17 [CD, VB.

[CD. VB, BYPASS, BYP~SS. RPV pressure <200 psig, psig. No CCI]

CCI]

Large break interfacing system LOCA outside containment occurs. Core damage damage (CD)

(CD) and subsequent vessel breach (VB) results at low RPV pressure with a a bypassed containment. There are no core concrete interactions (No CCI) CCI) due to the presence of an overlying pool of water.

E.1-47 E.1-47 { .:

"-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 Environmental Report Operating Operating License Renewal Stage E.1-9 Table E.1-9 Progression Bins (CAPB)

Collapsed Accident Progression (CAPB) Descriptions Descriptions (Continued)

(Continued)

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

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

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

CCI]

Large break interfacing system LOCA outside containment occurs. Core damage Core damage (CD) and subsequent vessel breach (VB)

(CD) (VB) results at low RPV pressure with a 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 summarized in Tables

%mv E.1-10 and E.1-11 respectively. Table E.1-10 summarizes the results of E.1-1 0 summarizes the CET ofthe 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.

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 Applicant's Environmental l *1'.

Operating License Renewal Stage (p-E.1-100 Table E~1*1 Summary of PNPS Containment Containment Event Tree Quantification Release Category Category Release Frequency (Timing/Magnitude)

(Timing/Magnitude) (/RY)

Late Low 4.53E-06 Medium Late Medium 1.56E-06 1.56E-06 Late High O.OOE-00 O.OOE-OO Early Low 3.32E-08 3.32E-OB Early Medium 6.48E-08 6.4BE-OB Early High 1.13E-07 1.13E-07 Containment Failure No Containment 1.11E-07 1.11E-07 Nomenclature Nomenclature Timing L (Late) - Greater E (Early)

E Greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Early)-- Less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> uJW)

Magnitude NCF NCF (Little to no release) 0.001% Csi

- Less than 0.001% Csl LO LO (Low) - 0.001 to 1 1% Cs1

% Csi MED (Medium)

MED (Medium) -1 to 10%

10% Csi Csl Hi HI (High) - Greater than 10% 10% Csl Csi E.1-49

("V

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~--------

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR PilgrimNuclear Pilgrim PowerStation NuclearPower Station Applicant's EnvironmentalReport Applicant'sEnvironmental Report OperatingLicense Operating RenewalStage LicenseRenewal Stage TableE.1-11 Table E.1-11 Collapsed ProgressionBin AccidentProgression CollapsedAccident (CAPB)Source Bin(CAPB) SourceTerms Terms '  !

CAPB CAPB Warning Warning Release Release Release Release Release Release Elevation Elevation

- CAPB CAPB Frequency Frequency Time Time Start Start I Duration Duration Energy Energy*

(m)

(m)

(Iyear)

(/year) (sec)

(sec) (sec)

(sec) (sec)

(sec) (W)

(W) 11 CAPB-1 CAPB-1 9.51 E-08 9.51E-OB 3.98E+03 3.9BE+03 3.OOE+01 3.00E+01 2.20E+04 2.20E+04 9.00E+03 2.61E+05 9.OOE+03 2.61E+05 22 CAPB-2 CAPB-2 1.27E-08 1.27E-OB 3.96E+03 3.96E+03 3.OOE+01 3.00E+01 2.20E+04 2.20E+04 9.00E+03 2.50E+05 9.OOE+03 2.50E+05 33 CAPB-3 CAP B-3 2.39E-09 2.39E-09 3.96E+03 3.96E+03 3.OOE+01 3.00E+01 2.20E+04 2.20E+04 9.OOE+03 2.50E+05 9.00E+03 *2.50E+05 44 CAPB-4 CAPB-4 3.29E-09 3.29E-09 7.96E+03 7.96E+03 3.OOE+01 3.00E+01 1.83E+04 1.B3E+04 3.56E+03 3.S6E+03 1.IOE+07 1.10E+07 5S CAPB-5 CAPB-S 2.73E-09 2.73E-09 1.31 E+04 1.31E+04 3.OOE+01 3.00E+01 2.53E+04 2.53E+04 7.93E+03 7.93E+03 8.34E+06 B.34E+06 66 CAPB-6 CAPB-6 7.95E-09 7.9SE-09 1.33E+04 1.33E+04 3.OOE+01 3.00E+01 2.56E+04 2.56E+04 8.11E+03 B.11E+03 8.23E+06 B.23E+06 77 CAPB-7 CAPB-7 7.93E-09 7.93E-09 1.38E+04 1.3BE+04 3.OOE+01 3.00E+01 2.61 E+04 2.61E+04 8.46E+03 B.46E+03 8.03E+06 B.03E+06 8B CAPB-8 CAPB-8 2.06E-08 2.06E-OB 9.18E+03 9.1BE+03 3.00E+01 3.00E+01 2.OOE+04 2.00E+04 4.59E+03 4.S9E+03 1.04E+07 1.04E+07 99 CAPB-9 CAPB-9 9.25E-09* 9.21 9.25E-09 E+03 9.21E+03 3.OOE+01 3.00E+01 2.44E+04 2.44E+04 8.87E+03 B.B7E+03 4.18E+06 4.1BE+06 10 10 CAPB-10 CAPB-10 8.53E-08 B.S3E-OB 1.37E+04 1.37E+04 3.OOE+01 3.00E+01 2.60E+04 2.60E+04 8.40E+03 B.40E+03 8.06E+06 B.06E+06 11 11 CAPB-11 CAPB-11 4.35E-08 4.35E-OB 1.37E+04 1.37E+04 3.OOE+01 3.00E+01 2.60E+04 2.60E+04 8.40E+03 B.40E+03 8.06E+06 B.06E+06 12 12 CAPB-12 CAPB-12 1.70E-06 1.70E-06 2.84E+04 2.B4E+04 3.OOE+01 3.00E+01 4.64E+04 4.S4E+04 9.OOE+03 9.00E+03 7.59E+06 7.S9E+06 13 13 CAPB-13 CAPB-13 2.30E-09 2.30E-09 9.14E+03 9.14E+03 3.OOE+01 3.00E+01 2.71E+04 ' 9.OOE+03 2.71E+04 9.00E+03 1.80E+06 1.BOE+06 14 14 CAPB-14 CAPB-14 2.26E-06 2.26E-06 2.66E+04 2.66E+04 3.OOE+01 3.00E+01 4.46E+04 4.46E+04 9.OOE+03 9.00E+03 7.08E+06 7.0BE+06 15 15 CAPB-15 CAPB-15 2.12E-06 2.12E-06* 2.81 E+04 2.B1E+04 3.OOE+01 3.00E+01 4.62E+04 4.62E+04 9.OOE+03 9.00E+03 7.60E+06 7.60E+06 16 16 CAPB-16 CAPB-16 1.18E-09 1.1BE-09 3.96E+03 3.96E+03 3.OOE+01 3.00E+01 2.12E+04 2.12E+04 9.OOE+03 9.00E+03 2.50E+05 2.50E+OS 17 17 CAPB-17 CAPB-17 6.91E-09 6.91E-09 3.96E+03 3.96E+03 3.OOE+01 3.00E+01 2.14E+04 2.141::+04 9.OOE+03 9.00E+03 2.50E+05 2.S0E+OS*

18 1B CAPB-18 CAPB-1B 4.61E-10 4.61E-10 3.96E+03 3.96E+03 3.OOE+01 3.00E+01 2.12E+04 2.12E+04 9.OOE+03 9.00E+03 2.50E+05 2.50E+05 19 19 CAPB-19 CAPB-19 2.43E-08 2.43E-OB 3.96E+03 3.96E+03 3.OOE+01 3.00E+01 2.18E+04 2.1BE+04 9.OOE+03 9.00E+03 2.50E+05 2.50E+OS E.1-50:

E.1-50

Exhibit No. NRC000001 NRC - Applicant's Environmental Report Pilgrim LR Proceeding SAMA Analysis 50-293-LR, 06-848-02-LR Pilgrim PilgrimNuclear NuclearPower PowerStation Station Applicant's Environmental Report Applicant's Environmental Operating License Operating Report Renewal Stage License Renewal Stage L')'

Table Table E.1-11 E.1-11 Collapsed Progression Bin Accident Progression Collapsed Accident Bin (CAPB) Source Terms (CAPB) Source Terms (continued)

(continued)

I Release Release Fractions Fractions NG NG  ! T Cs Cs Te Te Sr Sr Ru Ru La La Ce Ce Ba Ba 11 1.99E-07 1.85E-07 O.OOE+O0 1.85E-07 1.85E-07 1.99E-07 1.85E-07 8.00E-09 5.01E11 1.24E-09 8.OOE-09 O.OOE+OO 1.24E-09 8.43E-11 1.70E-08 5.01E.,.11 8.43E-11 1.70E-08 22 9.97E...QS 4.81 9.97E-05 4.81E...QS 3.97E-07 4.OOE-06 1.76E-07 3.97E-07 4.66E-OS 1.76E-07 E-05 4.66E-05 4.00E-06 *1.6SE...Q8 S.1SE-08 4.87E-06 1.65E-08 5.15E-08 4.87E-06 33 9.97E-05 4.97E-OS 1.76E-06 S.37E-OS 4.97E-05 9.97E-OS 5.37E-05 4.00E-06 2.37E-08 S.80E-07 4.OOE-06 1.76E-06 5.80E-07 1.S7E-07 4,95E-06 2.37E-08 1.57E-07 4.9SE-06 44 1.OOE+00 2 . 62E-02 4.18E-05 4.90E-02 2.62E-02 1.00E+00 4.90E-02 3.66E-04 ' 8.97E-07 2.46E-OS*. 3.66E-04 4.18E-OS 2.46E-05 3.04E-06 1.92E-04 8.97E-07 3.04E-06 1.92E-04 5S 9.85E-01 3.68E-02 4.28E-05 7.86E-02 3.68E-02 9.8SE-01 7.86E-02 3.66E-04 1.56E-06 4.10E-OS 3.66E-04 4.28E-05 4.1OE-05 6.79E-06 3.44E-04 1.S6E-06 6.79E-06 3.44E-04 66 1.OOE+00 2.32E-02 1.48E-03 4.02E-02 2.32E-02 1.00E+00 4.02E-02 3.66E-04 6.50E-06 3.19E-04 3.66E-04 1.48E-03 3.19E-04 7.17E-OS 3.23E-04 6.50E-06 7.17E-05 3.23E-04 77 9.76E-01 6.11 9.76E-01 E-02 2.94E-02 6.11E-02 3.66E-04 9.1 2.30E-04 3.66E-04 1.26E-03 2,30E-04 2.94E-02 1.26E-03 9.10E-06 1~06E-04 4.52E-04 OE-06 1.06E-04 4.52E-04 88 1.OOE+00 2.72E-01 3.07E-05 2.98E-01 2.72E-01 1.00E+00 2.98E-01 2.23E-02 4.49E-05 9.89E-04 2.23E-02 3.07E-05 9.89E-04 6.S7E-05 1.1 4.49E-OS 6.57E-05 5E-02 1.1SE-02 99 5.97E*01 7.61 5.97E-01 7.07E-02 1.41 E-02 7.07E-02 7.61E-02 1.41E-OS 9.72E-04 1.09E-02 E-05 9.72E-04 7.63E-OS 1.02E-02 3.69E-OS 7.63E-05 1.09E-02 3.69E-05 1.02E-02 10 10 1.OOE+00 2.49E-01 1.1 2.80E-01 2.49E-01 1.00E+00 2.80E-01 E-02 3.07E-03 1.11E-02 7.9SE-05 5.81 1.81E':02 7.95E-05 3.07E-03 1.81E-02 E-04 1.03E-02 S.81E-04 1.03E-02 (-j 11 11 9.79E-01 9.79E-01 1.73E-01 1.41 1.73E-01 E-01 9.97E-03 1.41E-01 3.13E-03 1.78E-02 9.97E-03 3.13E-03 9.39E-04 1.72E-02 1.22E-04 9.39E-04 1.78E-02 1.22E-04 1.72E-02 12 12 2.01 2.01E-01 5.84E-OS 4.37E-05 E-01 5.84E-05 2.36E-07 1.72E-06 1.25E-07 2.36E-07 4.37E-05 1.25E-07 2.56E-OB 2.99E-06 B.04E-09 2.56E-08 1.72E-06 8.04E-09 2.99E-06 13 13 7.99E-03 5.99E-03 9.97E-01 7.99E-03 9.97E-01 3.63E-OS 3.66E-04 1.76E-04 3.63E-05 5.99E-03 1.76E-04 2.15E-06 1.41 3.66E-04 2.15E-06 E-05 4.52E-04 1.41E-OS 4.52E-04 14 7.75E-01 2.67E-02 2.47E-05 2.8BE-02 2.67E-02 7.7SE-01 2.88E-02 2.13E-03 8.49E-06 2.0SE-04 2.13E-03 2.47E-05 2.05E-04 2.27E-05 2.61 8.49E-06 2.27E-05 E-03 2.61E-03 15 15 2.76E-01 2.68E-41 9.97E-01 2.76E-01 9.97E-01 2.27E-03 2.25E-02 1.27E-03 2.27E-03 2.6BE-01 1.27E-03 3.00E.,.04 2.74E-02 9.33E-05 3.OOE-04 2.25E-02 9.33E-05 2.74E-02 16 16 1.00E+00 6.71 1.OOE+00 6.71E-02 4.06E-04 9.11 3.26E-02 4.06E-04 E-02 3.26E-02 E-05 2.21 9.11E-05 2.21E-02 1.65E-OS 4.27E-05 1.45E-06 1.65E-05 E-02 1.45E-06 4.27E-OS 17 17 3.62E-01 3.37E-01 9.72E-01 3.62E-01 9.72E-01 3.37E-01 2.37E-03 2.20E-02 1.34E-03 2.37E-03 1.34E-03 1.62E-04 8.57E-03 9.90E-05 1.62E-04 2.20E-02 9.90E-05 8.S7E-03 18 18 9.76E-02 6.25E-02 1.00E+00 9.76E-02 1.OOE+00 4.67E-03 2.27E-02 2.09E-02 4.67E-03 6.2SE-02 2.09E-02 8.S0E-04 2.12E-03 7.4SE-05 8.50E-04 2.27E-02 7.45E-05 2.12E-03 19 19 3.77E-01 6.87E-02 4.03E-01 3.77E-01 9.72E...Q1 4.03E-41 9.72E-01 6.87E-02 9. 58 E-03 2.26E-02 9.58E-03 2.33E-03 1.20E-02 3.00E-04 2.33E-03 2.26E-02 3.OOE-04 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 Station Applicant's Environmental Environmental Report Operating license License Renewal Renewal Stage Stage E.1.2.2.7 Release Magnitude Calculations The MAAP MAAP computer code is used to assign both the radionuclide release magnitude magnitude and timing based on the accident progression characterization.

characterization. Specifically, MAAP MAAP provides the following information:

information:

• • containment containment pressure and temperature temperature versus time (time of containment failure is is determined by comparing these values with the nominal containment capability);

• • radionuclide release time and magnitude for a Jarge large number number of radioisotopes; and

• • release fractions for twelve radionuclide species.

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

A number of plant improvements improvements were identified in Table 2.4 of NUREG-1742, NUREG-1 742, Perspectives Gained from the IPEEE IPEEE Program, Program, Final Report, April 2002 [Reference E.1 -8]. These E.1-8].

improvements improvements were implemented.

The seismic CDF IPEEE was conservatively estimated to be 5.82x10-55 per reactor-year.

CDF in the IPEEE The seismic CDF CDF has recently been re-evaluated to reflect the updated Gothic computer code room heat up calculations that predict no room cooling requirements requirements for HPCI, HPCI, RCIC, RCIC, Core Spray, and RHR areas; to update random component failure probabilities; and to model model replacement of certain relays with a seismically rugged model. The updated updated seismic COF CDF of of 3.22x10-55 per per reactor-year was used in estimation of the factor of 6 used to determine the upper upper bound estimated benefit described in Section 4.21.5.4.

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

[Reference E.1-6]. The PNPS fire analysis was performed performed using the conservative EPRI's EPRl's Fire Induced Vulnerability Evaluation (FIVE) methodology for qualitative and quantitative screening of Evaluation (FIVE) of fire areas and for fire analysis of areas that did not screen [Reference E. E.11-7].

-71. The FIVE

. methodology is primarily a a screening approach used to identify plant vulnerabilities due to fire initiating events.

E.1-52 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 Applicant's Environmental Environmental Report Operating License Renewal Stage l,;i 0

Table E.1-12 presents the results of the PNPS IPEEE IPEEE fire analysis. The values presented presented in in Table E.1-12 E.1-12 are taken from NUREG-1742 NUREG-1742 [Reference E.1-8]. These values are the same as the

[Reference E.1-8].

IPEEE fire CDF original IPEEE CDF results (2.20E-5 (2.20E-5 per reactor-year) reactor-year) [Reference E.1-6] after the response to NRC NRC questions/issues regarding fire-modeling fire-modeling progression. A revised fire zone CDF of of 1.91 E-5 per reactor-year, generated generated to reflect updated equipment equipment failure probability and unavailability values was used in estimation of the factor of 6 used to determine determine the upper bound benefit described in Section 4.21.5.4.

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

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

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

events, examined hazards:

• • 0 high winds winds and tornadoes; tornadoes;

• • external flooding; and and

• • ice, hazardous chemical, transportation, and nearby facility incidents.

0 incidents.

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

forPNPS.

No risks to the plant occasioned by high winds and tornadoes, tornadoes, external floods, ice, Ice, and and transportation, and nearby facility incidents were identified that might lead hazardous chemical, transportation, lead damage with a predicted frequency in excess of 10-to core damage 1046/year.

/year. Therefore, these other other external event hazards are not included in this attachment and are expected riot not to to impact the conclusions of this SAMA SAMA evaluation.

E.1.4 PSA Model Review and Difference Peer Review Model Peer Current PSA Model between Current Difference between Model and 1995 1995 Update IPE Update E.1.4.1 PSA Model Model Peer Review Peer Review The original originallPEIPE PSA model was peer reviewed on March 2000 using the BWROG BWROG PSA Peer Peer Implementation Guidelines. Facts and Observation sheets documented Certification Implementation Review Certification documented the certification teams' insights and potential level of significance. As part of the update of the IPE PSA issues and observations from the BWROG models, all major issues PSA models, BWROG Peer Review (i.e., Level A, B, B, C,

C, and 0 observations) have been addressed and incorporated D observations) incorporated into the current IPE PSA model, April 2003 [Reference E.1-1]. E.1-1].

IPE/PSA model update, individual work packages (event tree, fault tree, human For the current IPE/PSA internal flooding analysis were circulated to each PSA reliability analysis (HRA), data, etc.) and !nternal PSA member for independent peer review. "The"accident member sequence packages, system work" The accident seqUence work " ; "

packages, HRA, and internal flooding analyses ~~ealsoassigned 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 department for review. Similarly, the accident sequence packages, system work packages, HRA report,!containment report, containment performance analysis, fault tree and event tree models, and Level 2 models were peer reviewed by an outside consultant.

project team and plant staff reviewed consequence and risk Entergy license renewal project The Entergy estimates for the SAMA 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 Pilgrim Nuclear Power Station Nuclear Power Applicant's Environmental Operating Environmental Report Operating License License Renewal Report Renewal Stage Stage U The peer peer review process process emphasized emphasized the role of plant staff, external consultants, consultants, and BWROG BWROG PSA PSA certification in this recent recent model update. The The peer reviews reviews served to ensure the accuracy of of both the assumptions made made in the models models and the results. The results of the peer peer review and and resolutions resolutions areare presented in Section 5 and Appendix P of the Pilgrim Nuclear Power Power Station Examination for Internal Individual Plant Examination Internal Events update update report, report, April 2003 [~eference E.1-1].

2003 [Reference E.1-1].

E.1.4.2 Major Differences E.1.4.2 Major Differences between between the Updated Updated IPE IPE PSA PSA Model Model and 1995 Update IPE 1995 Update IPE Model Model E.1.4.2.1 E.1.4.2.1 Core Damage Damage -- Comparison Comparison to the PNPS 1995 1995 Update Update IPE IPE Model Model The current PNPS IPE/PSAIPE/PSA update model model was completely revised in response to the BWROG BWROG Peer Review of March 2000 [Reference E.1-1]. The updated updated model is based upon all procedures and plant design as of September September 30, 2001, 2001, and plant data as of DecemberDecember 31, 2001. 2001. The results yield a measurably lower CDF CDF (point estimate CDF CDF - 6.41 E-6/reactor year)year) than the original IPE IPE (point estimate CDF - 5.85E-5/yr) estimate CDF 5.85E-5/yr) [Reference E.1-2] E.1-2] and 1995 1995 PSA model update (point estimate CDF 2.84E-5/yr) [Reference E.1-3].

CDF - 2.B4E-5/yr) E.1-31. (The performed to (The 1995 update was performed answer NRC questions following the IPE submittal.) The improved IPE submittal.) improved results are due to improvedimproved*

plant performance, performance, replacement switchyard -breakers, replacement of switchyard breakers, more realistic success criteria criteria based on MAAP runs, and more MAAP more sophisticated data handling.

handling. Major changes are summarized 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 originallPE 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 originallPE original IPE were completely revisedrevised... BWROG BWROG certification findings and incorporated into the revised event trees. Major facts and observations observations were incorporated include the following.

(1) LOOP Event Tree (1)

The LOOP event was completely revised to account for failure modes of HPCI/R.GIC 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; operation; RPV depressurization on HCTL; and transfer to the SBO SBO tree to address such items as premature battery depletion and AC recovery at 30 minutes and beyond.

E.1 C>

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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 Environmental Report Operating License Renewal Renewal Stage seo Event Tree (2) SBO (2)

Current update reflects GE load shed calculations and use of plant SBa SBO procedures for DC load shedding.

(3) Inadvertent Stuck Open Relief Valve (I0RV)

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

(4) LOCAs (4) LOGAs Event Trees

. The update considers both HPCI and RCIC RCIC for small break LOCAs.

LOCAs.

Large and medium LOCAs and subsequent ATWS are modeled as core damage damage end states in the updated model.

model. Small break LOCAs LOCAs and ATWS ATWS are treated as similar to transient-induced ATWS.

The vapor suppression suppressio", system is considered during during large LOCAs events.

(5)

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

The revised ATWS ATWS model takes into consideration "inhibit ADS" ADS" and MSIV bypass issues. In addition, HRA values take into consideration consideration ATWS accident progressions for RPV and containment conditions predicted by MAAP MAAR.

(6) Loss-of-Gontainment (6) Loss-of-Containment Heat Removal Sequences The revised event trees model the potential impact impact from containment venting on low-pressure pressure system operation. For examplE!,

example, no credit is given for core spray and LPCI if credi~ isgivel1 containment venting is required. In related phenomena, In addition, other containment relateCJ phenomena, such as high torus temperatures (HPCI) and high containment pressures (RCIC, temperatures (HPCI) (RClq. SRVs)

SRVs) are reflected in the updated event trees. .

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

(7) ISLOCA (7) ISLOGA Event Tree NSAC-154 [Reference E.1-10] and NUREG/CR-5124 NSAC-154 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 low-pressure injection 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 Environmental Report Operating license Report License Renewal Stage U The revised ISLOCA ISLOCA event tree credits use of condensate or fire water for large ISLOCA ISLOCA events provided that LPCI or core spray operation had previously occurred occurred to provide initial RPV reflood.

(8) Other Changes (8)

The revised event trees credit use of feedwater when appropriate.

Control 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 containment pressure with direct torus containment venting. -

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

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

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

temperature exceeds 2200°F.

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

The MAAP4 MAAP4 computer computer code [Reference E.1-4]

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. LOCAs.

• • Success criteria for SORV SORV are same as for non-SORV cases (2 SRVs are required for successful RPV depressurization).

• • Consistent success criteria are used for RPV depressurization for transients, medium medium LOCAs, LOCAs, and small LOCAs.

• • Plant specific calculations were performed performed to identify the plant response for single or double recirculation pump trip failures.

• • The appropriateness 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 Environmental Report Hi'" Operating License Renewal Stage Operating

• In addition to the MAAP4 code, the GOTHIC code [Reference In [Reference E.1-13]

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 D.

System fault tree models from the original IPE were completely revised to reflect the as-built System MAAP analyses were clearly identified to support the success criteria of plant configuration. MAAP 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 models. NUREG/CR-5500 [Reference E. 1-9], which decreased the failure-to-scram probability from 3.OE-5/yr E.1-9], 3.0E-5/yr to 5.8E-6/yr.

E. DataAnalysis E Data Component failure data, both generic and plant-specific, were reviewed and updated with more Component recent experience (the performance of risk significant Significant systems HPCI HPCI and RCIC has greatly greatly improved since the original IPE). Plant-specific data were adjusted for industry originaIIPE). industry experience using Maintenance unavailability values were updated based on maintenance rule Bayesian updates. Maintenance records from the system engineers. More recent common common cause failure data and appro~ch approach NUREG/CR-5497 [Reference E.1-14]

NUREG/CR-5497 E.1-14] were factored into this update. In In particular, a*more a more methodology (Alpha model) common-cause failure methodology detailed and refined common-cause model) has been applied in this this update.

update. In more common-cause In addition, more common-cause equipment failure groups such as fans, dampers, dampers,

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

F. HA F. HRA performed to identify, quantify, and document A complete revision of the HRA was performed document the pre-initiator and post-initiator human errors (including recoveries). The updated HRA was performed NUREG/CR-1278 [Reference E.1-15], also referred to as THERP.

using NUREG/CR-1278 THERP. Screening Screening values were only used for low-significance human errors. In low-significance human In addition, a detailed analysiS analysis was performed performed to between post-initiator treat dependencies between post-initiator errors.

Dependencv Analysis G Dependency A complete revision of the internal flooding analysis was developed to systematically address spatial spatial. dependencies. i;*  !

! i Dependency between pre-initiator human Dependency of instruments) was human errors (such as miscalibration oflinstrum~nts) modeled. In addition, dependencies between multiple post-accident operator.

modeled. operator actions appearing appearing in the same accident sequence were evaluated. - ,I i I:

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

E.1-58 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 Power Station Environmental Report Applicant's Environmental Operating License license Renewal Renewal Stage U H. Structural Response The ISLOCA frequency was revised.

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

1.

I. Quantification The truncation value was lowered to 1.0E-11.I.OE-11.

Human Error Probability (HEP)

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

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

The HRA was completely revised to address a comment completelY'revised comment from the PSA Certification Certification [Reference E.1-16]

E.1-16] that many of the HEPs were not realistic using the previous methodology. In many cases (e.g., failure to perform perform DTV),

DTV), the previous HEPs were judged to be overly conservative.

J. Internal Flooding Analvsis Analysis The internal flooding analysis from the original IPE systematic systematic examination of the flood source and IPE was completely revised to include a detailed, progression for each of the analyzed flooding U

scenarios. In addition, the updated internal internal flooding analysis considers the effects of spray on equipment.

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

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

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

• • CET fault models were revised to ensure mitigating systems were not degraded in ensure that mitigating in' the Level I1 sequence.

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

E.1-59 L.'i

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

• • Shell melt-through melt-through phenomena were considered where applicable. applicable .

• • 'Operator Operator responses to key actions were reassessed to incorporate incorporate the probability for success given the containment conditions and Emergency 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 E.1.S MACCS2 Model - levelLevel 3 Analysis E.1.5.1 Introduction E.1.S.1 SAMA SAMA evaluation relies on Level 3 PRA results to measure measure the effects of potential plant modifications. A Level 3 PRA model using the MACCS2 MACCS2 [Reference E.1-18] E.1-18] was created for PNPS. This model, which requires detailed site-specific meteorological, meteorological, population, and economic data, estimates the consequences in terms of population dose and offsite economic economic cost. Risks in terms of population dose risk (PDR) (PDR) and offsite economic cost risk (OEeR) were (OECR) 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 11 is the base base case with delayed evacuation. Sensitivity case 2 is is the base caS!3 case with lower evacuation speed.

" t PDR was estimated by summingsumming over all releases the product of population dose and frequency for each accidental release. Similarly, OECR OECRwas was estimated by summing summing over all releases the product of offsite economic cost and frequency for each accidental release. Offsite economic economic during :the cost includes costs that could be incurred dJHng the emergency emergency response phase and costs that could be incurred through long-term long-term protective actions.

E.1.5.2 Input E.1.S.2 Input

,I ~ j Ii j , ,

The following sections describe the site-speci~icinput site-specific ninput parameters used to obtain the off-site dose and economic impacts for cost-benefit analyses.

analyses, :, I! .

E.1~O 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 Power Station Operating license Report Applicant's Environmental Report License Renewal Stage Stage C,)

E.1.5.2.1 E.1.S.2.1 Projected Projected Total Population by Spatial Element Element The total population within a SO-mileradius 50-mile radius of PNPS was estimated for the year 2032, the end of of the proposed license renewal period, for each spatial element element by combining total resident population projections with transient population data obtained from Massachusetts.and Massachusetts and Rhode Rhode Island. Table E.1-13 E.1-13 shows the estimated population distribution.

Table E.1-13 Estimated Population Distribution within a 50-mile Radius 0-10 10-20 10-20 20-30 30-40 30-40 40-50 50-Mile Sector Sector Miles . Miles Miles Miles Miles Miles Total Total*

Miles Miles Miles Miles N

N  ; 0 0 0 0 0 80474 80474 NNE 3 0 0 0 0 3 NE 3 0 0 0 0 0 0 3 ENE 3 0 33121 0 0 0 33124 E 5 0 0 33121 23185 0 0 56311 ESE ESE 23 0 49682 92740 0 142445 SE 950 9936 115925 23185 0 149996 SSE 13289 69555 82803 0 0 165647 165647 Q.,

S 23695 99364 132485 132485 84383 43397 383324 SSW 23695 49762 23696 23185 23185 21699 21699 142037 142037 SW 23695 71088 277374 277374 349491 114546 a36194 836194 WSW WSW 23695 23695 71088 277374 277374 349491:<

349491, 183037 904685 W

W 22818 71088 277374 277374 388324 286370 286370 1045974 1045974 WNW 16494 71088 118481 303450 390150 899663 NW 11269 11269 71088 195075 1529212 405561 2212205 NNW 5599 35544 43350 31295 31295 321894 437682 437682 Total 165236 619601 1659861 1659861 3197941 3197941 1847128 7489767 The 2000 U.S. Census Bureau Bureau data, togetherwith together with Massachusetts and Rhode Island population population projection data, was used to project county-level resident populations to the year 2032.

transient/resident Seasonal peak transient population was conservatively used to establish a transient/resident population ratio for each county within the 50-mile 50-mile radius. The ratio was found to be decreasing decreasing over time. For purposes of this study, the total county level population values were estimated by E.1-61 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 Pilgrim PowerStation NuclearPower Station Applicant's EnvironmentalReport Applicant's Environmental Report Operating License Operating Renewal Stage LicenseRenewal Stage summing the summing the year 2000 peak year 2000 population of transient population peak transient of each county and each county projected year the projected and the year 2032 2032 population of permanent population permanent county to that county ofthat to obtain the 2032 obtain the county population.

total county 2032 total population.

E.1.5.2.2 E.1.5.2.2 Land Fraction Land Fraction '

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

E.1.5.2.3 E.1.S.2.3 Watershed Class Watershed Class are two There are There watershed types two watershed types in In the SO-mile zone the 50-mile surrounding PNPS:

zone surrounding ocean and PNPS: ocean and land land (watersheds) drained (watersheds) drained by by rivers. There are rivers. There are no lakes. The major lakes.

no major The watershed index assigns watershed index assigns "`" "1" to to any element having spatial element any spatial non-zero land having aa non-zero fraction and land fraction and "2""2" to to all elements over all elements the Atlantic over the 'Atlantic Ocean or Ocean or its its bays.

bays.

E.1.5.2.4 E.1.S.2.4 Economic Data Regional Economic Regional Data Region Index RegaLon Index Each spatial Each element was spatial element assigned to was assigned economic region, an economic to an defined in region, defined in this report as this report as aa county. Where a spatial county. Where portions of covers portions element covers spatial element than one more than of more county, itit was one county, was assigned to assigned to that county having that county most area having the most within the area within the element.

element.

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

Agriculture.

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

data for 2002.

VALWF --Value VALWF Value of of Fain, Farm Wealth Wealth, '

MACCS2 requires an MACCS2 value of farm average value an average (dollars/hectare) for wealth (dollars/hectare) farm wealth the 50-mile for the 50-mile radius area PNPS. The around PNPS.

area around county-level farmland The county-level value was property value farmland property used as was used as a a

basis deriving this value. VALWF basis for deriving VALWF is $23,578/hectare.

$23,S78/hectare. "

VALWNF- Value of Non-Farm VALWNF- Non-Farm Wealth MACCS2 also requires MACCS2 average value of non-farm requires an average wealth. The non-farm wealth. county-level non-The county-level non-farm property value was used as aa basis for farm deriving this value. VALWNF for. deriving VALWNF is is $189,041/

$189,0411 person.

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

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 i

~

Operating License license Renewal Stage I

)

Variable Variable Description Value EVACST EVACST Daily cost for aa person who has been evacuated evacuated 42.3

($/person-day)

($/person-day)

POPCST POPCST Population relocation cost ($/person)

Population ($/person) 7840 7840 RELCST RELCST Daily cost for aa person who is is relocated ($/person-day)

($/person-day) 42.3 CDFRMO CDFRMO decontamination for the various levels of Cost of farm decontamination 881

($/hectare) decontamination ($/hectare) 1959 CDNFRM CDNFRM Cost of non-farm non-farm decontamination decontamination for the various levels of 4700

($/person) decontamination ($/person) 12540 DLBCST DLBCST Average cost of decontamination labor ($/person-year)

($/person-year) 54800 DPRATE DPRATE Property depreciation rate (per year) 0.2 DSRATE DSRATE Investment rate of return Investment return (per year) year) 0.12 0.12 E.1.5.2.5 E.1.S.2.S Agriculture Data The source of regional crop information 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 MACCS2 crop categories.

E.1.5.2.6 E.1.S.2.6 Meteorological Data MACCS2 model requires meteorological data for wind speed, wind'direction, The MACCS2 wind direction, atmospheric atmospheric stability, accumulated precipitation, and atmospheric mixing heights. The required data was obtained from the PNPS site meteorological monitoringmonitoring system and the Automated Automated Surface Observatory System (ASOS)(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 the second located loCated west of the old l&S I&S building, the "lower" and "upper" "lower" "upper' towers respectively. The upper tower is the designateddesignated data source for MACCS2 MACCS2 input. Data from the lower tower was'used was* used only if measurements 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%

98% complete. Missing data was obtained either from the lower tower or from estimates based on adjacent valid measurements measurements of the missing hour.

E.1 u

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 Applicant's Environmental Report Operating License Renewal Renewal Stage Accumulated Accumulated Precipitation The nearest source of hourly precipitation data to PNPS is the ASOS at Plymouth Airport. The data was converted to MACCS2MACCS2 input format format to provide precipitation in hundredths hundredths of an inch.

Regional Mixing Height Data h~ight of the atmosphere above ground level within Mixing height is defined as the height will which a released contaminant will become mixed (from approximately (from turbulence) within approximately one hour. PNPS mixing Reference E.1-19, was used for MACCS2 mixing height data, given in Reference 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 E.1-20].

Emergency Plan [Reference E.1-20).

Emergency Evacuation Delay Time The elapsed time between siren alert and the beginning of evacuation is 40 minutes. 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 uncertainties in delay time.

in this study to evaluate consequence sensitivities due to uncertainties Speed Evacuation Speed The worst case for PNPS evacuation is during the winter, under adverse weather conditions, since snow removal can .add 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 movement 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 30 minutes, and the average evacuation speed miles/hour in clear weather ranges from 2.79 miles/hour miles/hour under adverse weather weather to 1.54 miles/houq.mder conditions. The average evacuation speed is 2.17 miles/hour, miles/hour, or 0.97 meter/second.

meter/second.

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

E.1.5.2.8 Inventory .

Core Inventory The estimated PNPS core inventory (Table E.1-14) E.1-14) used in the MACCS2MACCS2 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 NuclearPower Pilgrim Nuclear PowerStation Station Environmental Report Applicant's Environmental Applicant's Operating License Operating LicenseRenewal Report Renewal Stage Stage L.)

Table Table E.1-14 E.1-14 PNPS Core PNPS Inventory (Becquerels)

Core Inventory (Becquerels)

Nuclide Nuclide Inventory Inventory . Nuclide Nuclide Inventory Inventory Co-58 Co-S8 1.15E+16 1.1SE+16 Te-131m Te-131m 2.87E+17 2.87E+17 Co-60 C0-60 1.37E+16 1.37E+16 Te-132 Te-132 2.80E+18 2.80E+18 Kr-85 Kr-8S 1.88E+16 1.88E+16 1-131 1-131 1.94E+18 1.94E+18 Kr-85m Kr-8Sm 6.84E+17 6.84E+17 1-132 1-132 2.85E+18 2.8SE+18 Kr-87 Kr-87 1.24E+18 1.24E+18 1-133 1-133 4.07E+18 4.07E+18 Kr-88 Kr-88 1.68E+18 1.68E+18 1-134 1-134 4.45E+18 4.4SE+18 Rb-86 Rb-86 1.05E+15 1.0SE+15 1-135 1-135 3.83E+18 3.83E+18 Sr-89 Sr-89 2.08E+18 2.08E+18 Xe-133 Xe-133 4.07E+18 4.07E+18 Sr-90 Sr-90 1.47E+17 1.47E+17 Xe-135 Xe-13S 9.68E+17 9.68E+17 Sr-91 Sr-91 2.71E+18 2.71E+18

• Cs-134 Cs-134 3.17E+17 3.17E+17 Sr-92 Sr-92 2.83E+18 2.83E+18 Cs-136 Cs-136 8.51E+16 8.51E+16 Y-90 Y-90 1.58E+17 1.S8E+17 Cs-137 Cs-137 1.90E+17 1.90E+17 Y-91 Y-91 2.54E+18 2.54E+18 Ba-139 8a-139 3.75E+18 3.75E+18 Y-92 Y-92 2.84E+18 2.84E+18 Ba-140 Ba-140 3.70E+18 3.70E+18 Y-93 Y-93 3.23E+18 3.23E+18 La-140 La-140 3.77E+18 3.77E+18 Zr-95 3.34E+18 La-141 3.48E+18 Zr-95 Zr-97 Zr-97 Nb-95 3.34E+18 3.44E+18 3.44E+18 3.16E+18 La-141 La-142 La-142 Ce-141

.3.48E+18 3.35E+18 3.3SE+18 3.36E+18 U

Nb-95 3.16E+18 Ce-141 3.36E+18

. Mo-99 Mo-99 3.65E+18 3.65E+18 Ce-143 Ce-143 3.27E+18 3.27E+18 Tc-99m Tc-99m 3.15E+18 3.1SE+18 Ce-144 Ce-144 2.18E+18 2.18E+18 Ru-103 Ru-103 2.77E+18 2.77E+18 Pr-143 Pr-143 3.20E+18 3.20E+18

.Ru-105 Ru-10S 1.85E+18 1.8SE+18 Nd-147 Nd-147 1.43E+18 1.43E+18 Ru-106 7.52E+17 7.S2E+17

• Np-239 Np-239 4.26E+19 4.26E+19 Rh-105-Rh-10S 1.38E+18 1.38E+18

• Pu-238 Pu-238 2.96E+15 2.96E+1S Sb-127 Sb-127 1.74E+17 1.74E+17 Pu-239 Pu-239 7.51E+14 7.S1E+14 Sb-129 Sb-129 6.06E+17 6.06E+17 Pu-240 Pu-240 9.41 E+14 9.41E+14 Te-127 Te-127 1.69E+17 1.69E+17 Pu-241 Pu-241 1.62E+17 1.62E+17 Te-127m Te-127m 2.27E+16 2.27E+16 Am-241 Am-241 1.65E+14 1.6SE+14 Te-129 Te-129 5.68E+17 S.68E+17 Cm-242 Cm-242 4.35E+16 4.35E+16 Te-129m Te-129m 1.49E+17 1.49E+17 Cm-244 Cm-244 2.35E+15 2.3SE+1S Source: denved from Reference E.1-21 for a power level of 2028 MW(t)

Source: derived from Reference E.1-21 for a power level of 2028 MW(t)

E.1-65 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 Pilgrim PowerStation NuclearPower Station Applicant's Environmental Report Applicant's Environmental Report Operating License Operating RenewalStage LicenseRenewal Stage E.1.5.2.9 Source E.1.5.2.9 SourceTermsTerms Twelve release categories, Twelve release corresponding to categories, corresponding to internal sequences, were event sequences, internal event were part part of ofthe the MACCS2 input.

MACCS2 Details of input. Details ofthe termsfor sourceterms the source for postulated events are internal events postulated internal available in are available in on-site on-site documentation. A documentation. A linear release rate linear release was assumed rate was between the assumed between timethe the time release started the release started and and the the time time the release ended.

the release j

I'

. ended.

E.1.5.3 Results E.1.5.3 Results Risk estimates for Risk estimates one base for one case and base case two sensitivity and two cases were sensitivity cases analyzed with were analyzed MACCS2. The with MACCS2. The base case base assumes 40 case assumes delay and minute delay 40 minute and 0.97 speed of meter/secspeed 0.97 meter/sec evacuation. Sensitivity case of evacuation. Sensitivity case I1 the base isis the case with base case evacuation of delayed evacuation with delayed of22 hours. Sensitivity case hours. Sensitivity is the case 22 is base case the base with an case with an evacuation speed evacuation speed of 0.69 meter/sec.

of 0.69 meter/sec.

Table E.

Table E.1-15 shows estimated 1-15 shows base case estimated base mean risk case mean values for risk values release mode.

each release for each mode. The The estimated mean values of PDR and offsite OECR for PNPS are 13.6 person-rem/yr and estimated mean values of PDR and offsite OECR for PNPS are 13.6 person-rem/yr and

$45,900/yr, respectively.

$45,900/yr, respectively.

'*1 I' , I , i' E.1-66 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 Environmental Report Applicant's Environmental Report Operating License Renewal Renewal Stage \...)

Table E.1-15 E.1-15 Base Case Mean PDR and OECR OECR Values Pplto Population Offsite Offsite Offsite Offsite Population Dose Release Frequency Frequency Popsin Dose Economic Population Dose Economic Cost Cost Risk (PDR)

Mode Modoser (/yr) CstRisk Cost (PDR) . Rik(E Risk (OECR))

(person-sv)

(person-sv)1 Cs (person-rem/yr)

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

($) ($/yr)

CAPB-1 9.51 E-08 4.66E-01 3.82E+06 4.43E-06 2 CAPB-1 9.S1E-08 4.66E-01 3.B2E+06 4.43E-062 3.63E-01 CAPB-2 CAPB-2 1.27E-08 1.27E-OB 9.96E+01 9.96E+01 6.40E+06 1.26E-04 8.1OE-02 B.10E-02 CAPB-3 2.39E-09 1.06E+02 1.06E+02 6.48E+06 6.4BE+06 2.53E-05 2.S3E-OS .55E-02 11.SSE-02 CAPB-4 CAPB-4 3.29E-09 3.29E-09 1.38E+04 1.3BE+04 4.28E+09 4.2BE+09 4.54E-03 4.54E-03 1.41E+01 CAPB-5 CAPB-S 2.73E-09 2.73E-09 1.81E+04 1.B1E+04 5.30E+09 S.30E+09 4.94E-03 1.45E+01 1.4SE+01 CAPB-6 CAPB-6 7.95E-09 7.95E-09 1.51 E+04 1.S1E+04 3.51 E+09 3.S1E+09 1.20E-02 1.20E-02 2.79E+01 CAPB-7 CAPB-7 7.93E-09 7.93E-09 1.67E+04 1.67E+04 4.42E+09 4.42E+09 1.32E-02 1.32E-02 3.51 3.51E+01 E+01 CAPB-8 CAPB-B 2.06E-08 2.06E-OB 4.1OE+04 4.10E+04 1.47E+10 1.47E+10 8.44E-02 B.44E-02 3.03E+02 3.03E+02 CAPB-9 CAPB-9 9.25E-09 9.2SE-09 2.37E+04 2.37E+04 8.33E+09 B.33E+09 2.19E-02 2.19E-02 7.70E+01 CAPB-10 CAPB-10 8.53E-08 B.S3E-OB 4.31 E+04 4.31E+04 1.54E+10 1.54E+10 3.68E-01 3.6BE-01 1.31 E+03 1.31E+03 (w

CAPB-11 CAPB-11 4.35E-08 4.35E-OB 3.45E+04 3.4SE+04 1.15E+10 1.15E+10 1.50E-01 1.S0E-01 5.OOE+02 S.OOE+02 CAPB-12 CAPB-12 1.70E-06 1.70E-06 9.72E+01 9.72E+01 4.63E+06 4.63E+06 1.65E-02 1.6SE-02 7.88E+OO 7.BBE+OO CAPB-13 CAPB-13 2.30E-09 2.30E-09 7.30E+03 7.30E+03 6.53E+08 6.S3E+OB 1.68E-03 1.68E-03 1.50E+OO 1.S0E+OO CAPB-14 CAPB-14 2.26E-06 2.26E-06 1.58E+04 1.SBE+04 4.14E+09 3.57E+00 3.S7E+OO 9.36E+03 CAPB-15 CAPB-1S 2.12E-06 2.12E-06 4.31 E+04 4.31E+04 1.59E+10 1.S9E+10 9.14E+00 9.14E+OO 3.37E+04 3.37E+04 CAPB-16 CAPB-16 1.18E-09 1.1BE-09 1.86E+04 1.86E+04 5.50E+09 5.S0E+09 2.19E-03 2.19E-03 6.48E+OO CAPB-17 CAPB-17 6.91E-09 6.91E-09 4.81E+04 4.B1E+04 1.71E+10 1.71E+10 3.32E-02 1.18E+02 1.1BE+02 CAPB-18 CAPB-1B 4.61E-10 2.38E+04 2.3BE+04 7.86E+09 1.1OE-03 1.10E-03 3.62E+OO CAPB-19 CAPB-19 2.43E-08 2.43E-OB 5.31 E+04 5.31E+04 1.88E+10 1.BBE+10 1.29E-01 1.29E-01 4.56E+02 4.56E+02 Totals 1.36E+01 1.36E+01 4.59E+04 4.59E+04

1. 1

1. 1 sv=

sv = 100 rem 2.

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

(person-sv) x 100 (rern/sv)

(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 Operating Results of sensitivity analyses indicate that a delayed evacuation or a lower evacuation. evacuation speed would not have significant effects on the offsite consequences or risks determined in this study.

E.1-16 summarizes offsite consequences in terms of population dose (person-sv)

Table E.1-16 (person-sv) and .

offsite economic cost ($)

($) for the base case and the sensitivity cases. ComparisonComparison of the consequences indicates that the maximal d~viation deviation is less than 2% between the base case population dose and the Sensitivity Case 2 population dose for release mode CAPB-8. CAPB-8.

E.1-16 Table E.1-16 Summary of Offsite Consequence Sensitivity Results Population DoseDose (person-sv)

(person-sv) Offsite Economic Cost ($) ($)

Release 2-Hr 2-Hr Lower Lower 2-Hr 2-Hr Lower Lower Release Base Case Base Case Delayed Delayed Speed of of Base Case Delayed Speed ofof Mode Mode Evacuation . Evacuation Evacuation Evacuation Evacuation Evacuation Evacuation Evacuation CAPB-1 CAPB-1 4.66E-01 4.66E-01 4.66E-01 4.66E*01 4.67E-01 4.67E*01 3.82E+06 3.B2E+06 3.82E+06 3.B2E+06 3.82E+06 3.B2E+06 CAPB-2 CAP B-2 9.96E+01 9.96E+01 9.97E+01 9.97E+01 9.97E+01 9.97E+01 6.40E+06 6.40E+06 6.40E+06 CAPB-3 CAPB*3 1.06E+02 1.06E+02 1.06E+02 1.06E+02 1.06E+02 1.06E+02 6.48E+06 6.4BE+06 6.48E+06 6.4BE+06 6.48E+06 6.4BE+06 CAPB-4 CAPB-4 1.38E+04 1.3BE+04 1.39E+04 1.39E+04 1.39E+04 1.39E+04 4.28E+09 4.2BE+09 4.28E+09 4.2BE+09 4.28E+09 4.2BE+09 CAPB-5 CAPB*5 1.81 E+04 1.B1E+04 1.82E+04 1.B2E+04 1.82E+04 1.B2E+04 5.30E+09 5.30E+09 5.30E+09 5.30E+09 CAPB-6 CAPB*6 1.51E+04 1.51E+04 1.51E+04 1.51E+04 1.51E+04 1.51E+04 3.51E+09 3.51E+09 3.51E+09 3.51E+09 3.51E+09 CAPB-7 CAPB*7 1.67E+04 1.67E+04 1.68E+04 1:6BE+04 1.68E+04 1.6BE+04 4.42E+09 4.42E+09 4.42E+09 4.42E+09 4.42E+09 4.42E+09 CAPB-8 CAPB*B 4.1OE+04 4.10E+04 4.16E+04 4.16E+04 4.17E+04 4.17E+04 1.47E+10 1.47E+10 1.47E+10 1.47E+10 1.47E+10 1.47E+10 CAPB-9 CAPB-9 2.37E+04 2.37E+04 2.38E+04 2.3BE+04 2.39E+04 2.39E+04 8.33E+09 B.33E+09 8.33E+09 B.33E+09 8.33E+09 B.33E+09 CAPB-1D CAPB-10 4.31E+04 4.34E+04 4.34E+04 4.36E+04 4.36E+04 1.54E+1o 1.54E+10 1.54E+10 1.54E+10 1.54E+10 1.54E+10 CAPB-11 CAPB*11 3.45E+04 3.48E+04 3.4BE+04 3.49E+04 3.49E+04 1.15E+10 1.15E+10 1.15E+10 1.15E+10 1.15E+10 1.15E+10 CAPB-12 CAPB*12 9.72E+01 9.72E+01 9.75E+01 9.75E+01 ~.

9.78E+01 9.7BE+01 4.63E+06 4.63E+06 4.63E+06 4.63E+06 CAPB-13 CAPB-13 7.30E+03 7.30E+03 7.30E+03 7.31E+03 6.53E+08.

6.53E+08: 6.53E+08 6.53E+OB 6.53E+08 6.53E+OB CAPB-14 CAPB-14 1.58E+04 1.5BE+04 1.58E+04 1.5BE+04 1.59E+04 1.59E+04 4.14E+09 4.14E+09 4.14E+09 4.14E+09 CAPB-15 CAPB*15 4.31E+04 4.31E+04 4.33E+04 4.33E+04 4.35E+04 4.35E+04 1.59E+10 1.59E+10 1.59E+10 1.59E+10 1.59E+10 1.59E+10 CAPB-16 CAPB-16 1.86E+04 1.B6E+04 1.87E+04-1.B7E+04 1.88E+04 1.BBE+04 5.50E+09 5.50E+09 5.50E+09 5.50E+09 CAPB-17 CAPB-17 4.81E+04 4.B1E+04* 4.83E+04 4.B3E+04 4.86E+04 4.B6E+04 1.71E+10 1.71E+10 1.71E+10

. 1.71E+10 1.71E+10 1.71E+10 CAPB-18 CAPB*1B 2.38E+04 2.3BE+04 2.39E4-04 2.39E+04 2.40E+04 7.86E+09 7.B6E+09 7.86E+09 7.B6E+09 7.86E+09 7.B6E+09 CAPB-19 CAPB*19 5.31E+04 5.31E+04*. 5.33E+04

• 5.33E+04 5.37E+04 5.37E+04 1.88E+10 1.BBE+10 1.88E+10 1.8BE+10 1.88E+10 1.BBE+10 E.1-68 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 Power Station Operating Environmental Report Applicant's Environmental Operating License Renewal Stage Stage l....j E.1.6 References E.1-1 . ENN Engineering Report PNPS-PSA, "Pilgrim "Pilgrim Nuclear Power Station Individual Plant Examination for Internal Events Examination Events Update,"

Update," April 2003, Revision 1. 1.

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

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

letter dated December (2.95.127).

E.1-4 Program Boiling Water Reactor (MAAP Modular Accident Analysis Program (MAAP BWR) BWR) Code, Version 4.0.4 and Fauske & Associates, Inc., "MAAP "MAAP 4.0 Users manual,"

manual," prepared for The Electric Power Research Institute, May 1994.- 1994. .

E.1-5 E1-5 "The Implications Kaiser, nThe Implications of Reduced Source Terms for Ex-Plant Ex-Plant Consequence Modeling," Executive Conference on the Ramifications of the Source Term (Charleston, (Charleston, SC), March 12, 1985.

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 "ISLOCA Evaluation Guidelines,"

Guidelines,"

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

E.1-11 Chu, et al., "Interfacing Systems LOCA: Boiling Water Reactors,"

aI., "Interfacing Brookhaven National Reactors," Brookhaven National Laboratory, NUREG/CR-5124, BNL-NUREG-52141, February 1989.

NUREG/CR-5124, BNL-NUREG-52141, E.1-12 Electric Power Research Institute, "PSA "PSA Applications Guide, Guide,"H EPRI TR-1 05396, TR-105396, prepared by ERIN Engineering Engineering and Research, Inc., August, 1995.

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

E.1-69

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

Commission, NUREG/CR-5497, (INEEUEXT-97-01328),

Common-Cause Failure Parameter Common-Cause Estimations, October 1998.

Parameter Estimations, E.1-15 Swain, A. D.

D. and H. E.E. Guttmann, Guttmann, NUREG/CR-1 NUREG/CR-1278, 278, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications, Sandia National National Laboratories, U.S. Nuclear Regulatory Commission, Commission, August 1983.

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

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

E.1-18 E.1-18 Chanin, D. L. Young, Code Manual for MACCS2:

I., and M. l.

D. I., MACCS2: Volume 1, 1, User's User's Guide, SAND97-0594 SAND97-0594 Sandia National Laboratories, Laboratories, Albuquerque, NM, 1997. 1997.

E.1-19 E.1-19 Boston Edison Company, Company, "Appendix I Evaluation,"

Evaluation," forwarding evaluation of Pilgrim Station Unit 11 Conformance Conformance to the Design Objectives of 10 CFR 50, Appendix I, I, letter letter dated March 31, 1977 (2.77.031).

31,1977 (2.77.031).

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

November 1998.

\.~

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

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 PilgrimNuclear Pilgrim PowerStation NuclearPower Station Applicant's EnvironmentalReport Applicant's Environmental Report OperatingLicense Operating RenewalStage LicenseRenewal Stage ~,

ATTACHMENT E.2 ATTACHMENT E.2 SAMA CANDIDATES SAMA SCREENING AND CANDIDATES SCREENING AND EVALUATION 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 Environmental Report Report Operating License Renewal Stage E.2 EVALUATION EVALUATION OF SAMA SAMA CANDIDATES CANDIDATES This section describes the generation of the initial list of potential SAMA SAMA candidates, screening methods, and the analysis of the remaining SAMA SAMA candidates.

E.2.1 SAMA Compilation SAMA List Compilation A list of SAMA candidates was developeddeveloped by reviewing industry documentsdocuments 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 SAMA candidates from SAMA analyses for other GE plants. Industry documents documents reviewed include the following:

• • (Reference E.2-1),

SAMA Analysis (Reference Hatch SAMA

• • SAMA Analysis (Reference Calvert Cliffs Nuclear Power Plant SAMA E.2-2),

(Reference E.2-2),

• • GE ABWR ABWR SAMDA (Reference E.2-3),

SAMOA Analysis (Reference E.2-3),

• • Peach Bottom SAMA (Reference E.2-4),

SAMA Analysis (Reference E.2-4),

• • Quad Cities SAMA E.2-5),

SAMA Analysis (Reference E.2-5),

• • Dresden SAMA (Reference E.2-6), and SAMA Analysis (Reference

• • Arkansas Nuclear Unit 2 SAMA Evaluation Evaluation (Reference E.2-7). E.2-7).

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

documents. These sources of other industrydocuments

• • SAMOA cost estimate report (R~ference Limerick SAMDA (Reference E.2-8),

• • NUREG-1437 NUREG-1437 description of Limerick SAMDA SAMOA (Reference E.2-9), E.2-9),

• • NUREG-1437 NUREG-1437 description of Comanche Peak SAMOA of Comanche SAMDA (Reference E.2-10),E.2-1 0),

• • SAMDA submittal (Reference E.2-11),

Watts Bar SAMOA E.2-11),

• • TVA's NRC's RAI on the Watts Bar SAMDA TVA's response to NRC's (Reference E.2~

SAMOA submittal (Reference E.2-12),

12), Ii,

• • Westinghouse Westinghouse AP600 SAMDA (Reference E.2-13),

..* NUREG-0498, SAMOA(Reference NUREG-0498, Watts Bar Final Environmental Environmental Statement Supplement 1, 1, Section 7 7 (Reference E.2-14),

E.2-14),

• • NUREG-1 560, Volume 2, NRC Perspectives on the IPE Program NUREG-1560, Program (Reference (Reference E.2-15), .

and

• • NUREG/CR-5474, NUREG/CR-5474, Assessment of Candidate Candidate Accident Management Management Strategies (Reference (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 Nuclear Power Station Applicant's Environmental Environmental Report Operating License Renewal Stage L/

In addition to SAMA In SAMA candidates from review of industry documents, additional SAMA SAMA candidates were obtained from plant-specific sources, such as the PNPS IPE (Reference E.2-17)

IPE (Reference E.2-17) and IPEEE (Reference E.2-18). In both the IPE and IPEEE, IPEEE (Reference IPEEE, several enhancements enhancements related to severe accident insights were recommended and implemented. These enhancements enhancements are included in the comprehensive list of phase I SAMA SAMA candidates as numbers numbers 248 through 281. 281. The current 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 compret)ensive model were reviewed for similar failure modes and effects that could be addressed through a potential potential enhancement to the plant. The correlation enhancement between SAMAs correlation between SAMAs and the risk-significant terms were listed in Table E.1-2. .

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

SAMA E.2.2 Screenina of SAMA Qualitative Screening SAMA Candidates (Phase I)

The purpose of the preliminary SAMA screening was to eliminate from further consideration preliminary SAMA enhancements enhancements that were not viable for implementation implementation at PNPS. Potential SAMA SAMA candidates were screened out if they modified features not applicable to PNPS, if they had already been 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 SAMA SAMA candidate. During During this process, 63 of the phase I SAMA SAMA candidates were screened out because they were not applicable to PNPS, 4 of the phase I SAMA SAMA candidates were screened out because they were similar in nature and could be combined with another SAMA SAMA candidate, and 155 of the phase I U

SAMA candidates were screened out because they had already been implemented SAMA at PNPS, implemented atPNPS, leaving 59 SAMA candidates for further analysis. The final screening process involved identifying and eliminating those items whose implementation implementation .cost cost would exceed their benefit as described below. Table E.2-1 provides a description of each of the 59 phase 11 II SAMA SAMA candidates.

E.2.3 E.2.3 Final Screening and Cost Benefit Evaluation of SAMA Candidates Candidates (Phase (Phase IIIII)

A cost/benefit analysis was performed on each of the remaining remaining SAM SAMA A candidates.

candidates. If the implementation a implementation cost of a SAMA candidate was determined to be greater than the potential potential benefit (i.e.

(Le. 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 implementation of each SAMA SAMA was established from existing estimates of of similar modifications. Most of the cost estimates were developed from similar modifications considered in previously performed performed SAMA SAMA and SAMDA SAMOA analyses. In In particular, particular, these cost-estimates were derived from the following major sources:

• • GE ABWR SAMDA SAMOA Analysis (Reference E.2-3),

• • Peach Bottom Bottom SAMA SAMA Analysis (Reference E.2-4), E.2-4),

E.2-2 E.2-2 u

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

• • Quad Cities SAMA (Reference E.2-5),

SAMA Analysis (Reference E.2-5),

• • Dresden SAMA Analysis (Reference Dresden SAMA E.2-6),

(Reference E.2-6),

• • ANO-2 (Reference E.2-7),

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

deveioped for SAMDA costs were originally developed during the design phase of the plant),

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

contaminated material, etc.). Therefore, SAMA candidate was estimated in terms of averted implementing a SAMA The benefit of implementing consequences. The benefit was estimated by calculating the arithmetic difference difference between the four impact areas for the baseline plant design ,and associated with the four estimated costs associated total estimated and the total estimated impact area costs for the enhanced plant design (following implementationimplementation of the candidate).

SAMA candidate).

Values for avoided public and occupational health risk were converted to a monetary equivalent equivalent NUREG/BR-0184 (Reference E.2-19) conversion factor of $2,000' (dollars) via application of the NUREG/BR-0184.(Reference $2,000 per person rem and discounted to present value. Values for avoided off-site economic costs were also discounted to present present value. '

enhancement focuses> on establishing the economic viability of potential plant enhancement As this analysis focuses compared to attainable benefit, detailed cost estimates often were not required to make when compared modification. Several of the informed decisions regarding the economic viability of a particular modification.

informed SAMA candidates were Clearly SAMA clearly in excess of the attainable benefit estimated from aa 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 aa more detailed cost estimate was necessary to formulate a conclusion regarding the economic SAMA evaluations and previous submittals' SAMA viability of a particular SAMA. Based on a review of previous.submittals' implementation costs at PNPS, the following estimated costs for each an evaluation of expected implementation potential element of the proposed SAMA SAMA implementation implementation are ,used. used.

E.2-3 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 Environmental Report Operating License Renewal Report Renewal Stage L.)

Type of Change Estimated Estimated Cost Range Procedural only $25K-$50K

$25K-$50K Procedural change with engineering $50K-$200K

$50K-$200K*

required Procedural change with engineering engineering and $200K-$300K

$200K-$300K testing/training required Hardware modification $1OOK to >$1OOOK

$100K >$1000K In most cases, more detailed cost estimates were not required, particularly if the SAMA SAMA called for implementation of a hardware modification. Nonetheless, the cost of each unscreened the implementation unscreened SAMA SAMA candidate was conceptually estimated estimated to the pOint point where conclusions regarding the economic viability of the proposed modification could be adequately gauged. The cost benefit benefit comparison and disposition of each of the 59 phase "11 SAMA SAMA candidates is presented in Table E.2-1.

E.2-1.

Bounding evaluations (or analysis cases) were performed Bounding performed to address specific SAMA SAMA candidates or groups of similar SAMA SAMA candidates. These analysis cases overestimated overestimated the benefit and thus were conservative calculations. For example, one SAMA candidate suggested installing a digital C) large break LOCA LOCA protection system. The bounding calculation estimated estimated the benefit of this improvement by total elimination of risk due to large break LOCA improvement LOCA (see analysis in phase II11SAMA SAMA 052 of Table E.2-1). This calculation obviously overestimated overestimated the benefit, but if the inflated SAMA candidate was not cost beneficial, then the purpose of the benefit indicated that the SAMA analysis was satisfied.

A description of the analysis cases used in the evaluation follows.

Decay Capability - Torus Cooling Decav Heat Removal Capabilitv This analysis case was used to evaluate the change in plant risk from from installing an additional additional decay heat removal system. Enhancements Enhancements of decay heat removal capability decrease the probability of loss of containment heat removal. A A bounding analysis was performed performed by setting the events for loss of the torus cooling mode 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 II 11SAMAs SAMAs 1 1 and 14.

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

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

Filtered Vent Filtered Vent This analysis case was used to evaluate evaluate thethe change installing aa filtered risk from installing change in plant risk containment vent containment vent to provide fission product scrubbing.

to provide analysis was bounding analysis scrubbing. A bounding performed by was performed by reducing the successful torus torus venting accident progression venting accident terms by aa factor of 22 to source terms progression source reflect the reflect Reducing the additional filtered capability. Reducing the additional from the the releases from resulted in no the vent path resulted no benefit. This analysis case benefit. case was used to model model the benefit of phasephase 11 SAMAs 2 and 19.

II SAMAs 19.

Containment Vent for ATWS Containment Decav Heat Removal ATWS Decay Removal This analysis case was used to evaluate evaluate the change in plant risk from installing a containment containment alternate decay heat removal vent to provide alternate removal capability during an ATWS ATWS event. A bounding bounding performed by setting the ATWS analysis was performed sequences associated with containment bypass to ATWS sequences zero in the level I1 PSA PSA model, which resultedresulted in an an upper bound benefit of approximately approximately

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

of phase" SAMAs 47.

Core Debris Removal Molten Core Removal

~.; This analysis case was used to estimate the change in plant risk from providing a molten molten core debris cooling mechanism. A bounding analysis was performed setting containment 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 11 SAMAs 4, 5, 8, and 23.

II SAMAs Drywell Head Flooding Dryweff 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 temperature occurred, the drywell head seal 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 upper bound benefit of approximately $12,915. This analysis case was used to model the benefit of of phase 11SAMAs 6,18, phase" 6,18, and 20 20...

Reactor Building Effectiveness Reactor This analysis case was used used to evaluate the change in plant risk by ensuring the reactor reactor building is is available to provide effective fission product product removal.

removal. Reactor building building effectiveness was modeled by assuming reactor conservatively modeled reactor building availability for all accident sequences. This This resulted in resulted in an upper bound benefitbenefit of approximately $64,577. $64,577. This analysis case Vilas was used to to model the benefit model phase II SAMAs benefit of phase" SAMAs 7, 13, 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 Applicant's Report Environmental Report t:,

Operating License Renewal Operating Renewal Stage Stage """"

Strengthen Containment Containment This analysis case was used to evaluate the change in plant risk from strengthening containment containment to reduce the probability of containment containment over-pressurization over-pressurization failure. A bounding analysis was performed by setting all energetic containment containment failure modes (DCH, (DCH, steam explosions, late over-pressurization) to zero in the level 2 PSA model, which resulted in an upper bound benefit of of approximately $1,233,428. This analysis case was used to model the benefit of phase II11SAMAs .

10, 10, 15, 16, 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 melt-through does not occur. A bounding analysis was performed by setting containment failure due to base mat melt-through melt-through to zero in the level 2 PSA PSA model, which resulted in an upper bound benefit of approximately approximately $25,831. This analysis case was used to model the benefit of phase II11SAMASAMA 11. 11.

Coolinm Reactor Vessel Exterior Cooling This analysis case was used to evaluate the change in plant risk from from providing a method method to perform perform ex-vessel cooling of the lower reactor vessel head. A bounding bounding analysis was performed 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 II11SAMA SAMA 12.

Vacuum Breakers This analysis case was used to evaluate the change in plant risk from improving improving the reliability *of of vacuum breakers to reseat following a successful opening and eliminate suppression pool pool scrubbing failures from the containment analysis. A bounding analysis was performed 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 II11SAMA SAMA 17.

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

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

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 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 enhancement would extend HPCI and RCIC operability and allow more credit for AC power recovery. A bounding performed by changing the time available to recover offsite power before HPCI and analysis was performed 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 in the level 1I PSA model. This resulted in an upper bound benefit of approximately $146,356. This analysis case was used to model the benefit of phase" phase 11SAMAs 25, 26, 28, 33, and 35.

Improve DC Svstem System This analysis case was used to evaluate the change in plant risk from improving injection capability by auto-transfer auto-transfer of AC bus control power to a standby DC power source upon loss of of normal DC source or from enhancing procedure to make use of DC bus cross.:.tie the normal cross-tie to improve DC power availability and reliability. A bounding analysis was performed performed by setting the DC buses D1 6 and D1 bus'es 016 0177 to zero in the level 11 PSA model, which resulted in an upper bound benefit of of approximately $118,568. This analysis case was used to model the benefit of phase" approximately phase 11SAMAs 27 and 34. '

Altemate Pump Power Source Alternate This analysis case was used to evaluate the change in plant risk from adding a 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 performed by setting failure of of the SBO diesel generator to zero in level 1 1 PSA model, which resulted in in an upper bound benefit approximately $265,687. This analysis case was used to model the benefit of phase" of approximately phase 11SAMA SAMA 29.

System Improve AC Power Svstem was used to evaluate the change in plant risk from improving This analysis case was improving AC power system cross-tie capability to enhance the availability and reliability of the AC power system. A A bounding analysis was performed by setting the loss of MCCs B17, B18, and 815 B15 to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately approximately $473,410. This analysis case was used to model the benefit of phase phase"11SAMA 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 that would provide motive power to components (e.g., providing a dedicated DC DC power supply, additional batteries, or additional divisions). A bounding analysiS analysis was performed performed by setting the loss of DC bus D17 017 initiator, and one division of DC power, to zero in the level 1 1 PSA model, In an upper bound benefit of approximately which resulted In approximately $903,025. This analysis case was phase 11SAMAs used to model the benefit of phase" SAMAs 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 Environmental Report Operating License Renewal Stage ~)

Containment 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 ISLOCA event outside containment. A bounding analysis was performed by setting the RHR ISLOCA perfor/ned .by ISLOCA sequences to zero in the level 1 1 PSA model, approximately $16,497. This analysis case was which resulted in an upper bound benefit of approximately used to model the benefit of phase II11SAMA SAMA 36.

ISLOCA This analysis case was used to evaluate the change in plant risk from 'reducing reducing the probability of an ISLOCA ISLOCA by increasing the frequency of valve leak testing. A bounding analysis was performed by setting the ISLOCA performed ISLOCA initiator to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately approximately $24,148. This analysis case was used to model the benefit of phase II11SAMA SAMA 37.

MSIV Design MSIVDesign This analysis case was used to evaluate the change in plant risk from improving MSIV MSIV design to decrease the likelihood of containment bypass scenarios. A bounding analysis was performed performed by setting the containment bypass failure due to MSIV 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 II11SAMA SAMA 38.

Diesel to CST CST Makeup Pumps u

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

swapping HPCII RCIC suction from the CST to the torus allows continued HPCI and RCIC injection. Tt)erefore, Therefore, a bounding analysis was performed performed by setting the failure to switchover from CST to torus to zero in the level 11 PSA model, which resulted in no benefit. This analysis case was used to model the benefit of phase II11SAMA 39.

High Pressure Injection Iniection 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 powered high pressure injection system). A bounding analysis was performed performed by setting the CDF CDF contribution due to unavailability of the HPCI system to zero in the level 1 1 PSA PSA model, which resulted in an upper bound benefit of approximately $110,212. This analysis case was was used to model the benefit of phase IIII SAMAs 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 Power Station Applicant's Environmental Report Applicant's Environmental Report Operating License Renewal Stage Operating License Stage Reliabilitv of High Pressure Injection System Improve the Reliability Improve This analysis case was used to evaluate the change in plant risk from plant modifications modifications that that pressure injection system.

increase the reliability of the high 'pressure would increase bounding analysis was system. A bounding reducing the HPCI performed by reducing performed probability by a factor of three in the level 1 PSA HPCI system failure probability PSA model, which resulted in an upper model, upper bound approximately $76,025. This bound benefit of approximately This analysis case model the benefit of phase II11 was used to model SAMA 43.

SAMA '

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

$63,599. This analysis case was used to model the benefit of phase II11SAMA SAMA 46.

Diversity of Explosive Valves This analysis case was used to evaluate the change in plant risk from providing providing an alternate means of opening a pathway to the RPV for SLC system injection, thereby improving improving success probability for reactor shutdown. A bounding analysis was performed common cause performed by setting common SLG explosive valves to zero in the level 1 failure of SLC 1 PSA model, which resulted in an upper upper bound benefit of approximately $12,915. This analysis case was used to model the benefit of phase IIII SAMA 48.

SAMA48.

Reliability of SRVs This analysis case was used to evaluate the change in plant risk from installing additional signals improvement would reduce the likelihood of SRVs failing to automatically open the SRVs. This improvement to open, thereby reducing the consequences of medium lOCAs. 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 11 PSA model, which resulted in an upper upper bound benefit of approximately $31,799. This analySiS analysis case was used to model the benefit of phase II11SAMA 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 performed by depressurization to zero in the level 11 setting the probability of SRVs failing to open during RPV depressurization bound benefit model, which resulted in an upper bound PSA model, $194,378. This analysis benefit of approximately $194,378.

benefit of phase II11SAMA case was used to model the benefit SAMA 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 Power Station Applicant's Applicant's Environmental Environmental Report Operating License Renewal Stage ~)

Self-Cooled ECCS Pump Seals Self-Cooled ECCS 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 ECCS bounding analysis was performed by setting the CDF contribution from sequences involving RHR pump failures to zero in the level*

level 1 PSA model, which resulted in an upper bound benefit of of approximately $29,412. This analysis case was used to model the benefit of phase II11SAMA SAMA51. 51.

Large Break LOCA 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 11 PSA model, which resulted in an upper bound benefit of of approximately approximately $14,109. This analysis case was used to model the benefit of phase II11SAMA SAMA 52.

Containment Venting Controlled Containment 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 narrow pressure control band. This would prevent rapid containment depressurization when venting, thus avoiding adverse impact on the ability of the low pressure ECCS ECCS injection systems to take suction from the torus. A bounding performed by reducing the probability of the operator analysis was performed operator failing to recognize the need to vent the torus by a factor of three in the level 1 PSA model, model, which resulted in an upper bound benefit of approximately $137,237. This analysis case was used to model the benefit of phase II11 L)

SAMA SAMA53. 53.

ECCS ECCS Low Pressure Interlock '

This analysis case was used to* to evaluate the change in plant risk from installing a bypass switch to allow operator operator to bypass the ECCS low pressure interlock circuitry that inhibits opening of the RHR low pressure pressure injection and core spray injection valves following sensor or logic failure. A bounding analysis was performed performed by setting the CDF CDF contribution due to sensor failure, low pressure permissive logic failure, and miscalibration miscalibration to zero in the level 1 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 II11SAMA SAMA 54.

Improve the Reliabilitv Reliability of SSW and RBCCW RBCCW Pumps This analysis case was used to evaluate the change in plant risk from providing a separate pump pump train to eliminate common common cause failure of SSW and RBCCW RBCCW pumps. pumps. A bounding analysis was performed by setting the CDF performed CDF contribution due to common cause failures of SSW and RBCCW 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 II11SAMA SAMA 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 Environmental Report Applicant's Environmental Operating license License Renewal Stage Redundant DC Power Supplies to DTV Valve~ Valves This analysis casewas case was used to evaluate the change in plant risk from installing additional fuses fuses' DTV valve control circuits to enable the DTV function. A bounding analysis was performed to two DTV performed CDF contribution due to DC power supply failures to DTV by setting the CDFcontribution AO-5042B and DTV valves AO-5042B AO-5025 to zero in the level Jevel 1 PSA model. This resulted in an upper bound benefit of approximately $220,639. '-This This analysis case was used to model the benefit of phase" phase 11 SAMA 56.

SAMA Hydroturbine Proceduralize the Use of Diesel Fire Pump Hvdroturbine This analysis case was used to evaluate the change in plant risk from revising the procedure to hydroturbine allow use of hydro X-107A or diesel driven fire water pump P-140 is unavailable. A if EDG X-107A turbine ifEDG A bounding analysis was performed performed by setting the CDF CDF contribution from the sequences involving a LOOP and failure of either EDGEDG A or fuel oil (P-141) to zero in the level 1I PSA 011 transfer oil pump (P-141) PSA model. This resulted in an upper bound benefit of approximately approximately $175,279. This analysis case was used to model the benefit of phase "11SAMA 57.

Proceduralize Alignment Alignment of Bus B3 to Feed Bus BI B1 Loads or Bus B4 to Bus B2 This analysis case was used to evaluate the change in plant risk from providing a procedure to 480V MCCs direct the operator to restore 4BOV MCCs B B1515 and B B17 17 loads upon loss of 4.16kV bus A5 provided that 4.16kV bus A3 is available. The same is true for restoring 480V MCCs B 4BOV MCCs B14 14 and B18 B 18 loads upon loss of 4.16kV bus A6 provided that 4.16kV bus A4 is available. A bounding analysis was performed performed by setting the CDFCDF contribution from the sequences involving a loss of of

I';

, I the 4.16 kV bus A5 to zero in the level 1 PSA model. This resulted in an upper bound benefit of i:,; approximately $190,797. This analysis case was used to model the benefit of phase" phase 11SAMA SAMA 58.

Redundant 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 performed by setting the CDF contribution from the sequences involving fire water into LPCI LPCI loops A and B B cross-tie failure to zero in the level 11 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 II SAMA 59.

SAMA59.

E.2.4 Sensitivity Analyses E.2.4 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 Pilgrim Nuclear Power Station Applicant's Environmental Report {:,

Operating License Renewal Stage \...I Sensitivity Case 1: Years Remaining Until End of Plant Life Life The purpose of this sensitivity case was to investigate the sensitivity of assuming a 27 27-year

-year period for remaining plant life (Le.

(i.e. seven years on the original plant license plus the 20-year 20-year license renewal period was used in the base case. The license renewal period). The 20-year 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 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%

7.0% used in the base case analyses is conservative relative to corporate practices. Nonetheless, Nonetheless, a lower discount rate of 3.0% was assumed in this case to investigate the impact on each analysis case. ChangingChanging this assumption does not cause any any additional SAMAs SAMAs to be cost-beneficial.

E.2-12 E.2-12 u

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

the Edwin I. Hatch Nuclear Power Hatch Nuclear Plant Units Power Plant and 2, Units 11 and 2, March March 2000.

2000.

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

Supplement 1, February 1999.

E.2-3 E.2-3 General Nuclear Energy, Electric Nuclear General Electric Document for Support Document Technical Support Energy, Technical for the the ABWR, ABWR, 2SAS680, Revision 25A5680, January 18,1995.

Revision 1, January 18, 1995.

E.24 E.2-4 Appendix E-Appendix Environmental Report, E- Environmental Appendix G, Report, Appendix G, Severe Accident Mitigation Severe Accident Mitigation Alternatives Submittal Alternatives Related to Submittal Related to Licensing Renewal for Licensing Renewal for the Bottom Nuclear Peach Bottom the Peach Nuclear Power Plant Units 2 and 3, July, Power Plant Units 2 and 3, July, 2001. 2001.

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

2003.

E.2-6 E.2-6 Appendix AppendixF, Severe Accident F,Severe Alternatives Analysis Mitigation Alternatives Accident Mitigation Related to Submittal Related Analysis Submittal to Renewal for the Dresden licensing Renewal Licensing Nuclear Power Dresden Nuclear Units 22 and Plant Units Power Plant and 3, January 2003.

3, January 2003.

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

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

1989.

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

1996.

E.2-10 U.S.

E.2-10 Nuclear Regulatory U.S. Nuclear NUREG-1437, Generic Commission, NUREG-1437, Regulatory Commission, Environmental Impact Generic Environmental Impact Statement for License Renewal Statement Volume 1, Plants, Volume Nuclear Plants, Renewal of Nuclear Listing of S.36, Listing 1, 5.36, of SAMDAs SAMDAs considered for considered Comanche Peak for the Comanche Station, May Electric Station, Steam Electric Peak Steam May 1996.

1996.

E.2-11 E.2-11 Museler, W. J.,

Museler, (Tennessee Valley J., (Tennessee Authority) to Valley Authority) Document Control NRC Document to NRC Desk, "Watts Control Desk, "Watts Bar Bar Nuclear Plant (WBN)

Plant (WBN) Units I1 and 2 - Severe Accident Mitigation Design 2- Severe Accident Mitigation Design Alternatives Alternatives (SAMDAs),"

(SAMDAs),"letterletter dated October 7, dated October 1994.

7,1994.

E.2-12 Nunn, D. E.,

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

Nuclear Plant Bar Nuclear (WBN)

Units I1 and Units Accident Mitigation Severe Accident and 22 -- Severe Alternatives (SAMDA)

Design Alternatives Mitigation Design Response to (SAMDA) -- Response to Request for Additional Request (RAI) -- (TAC Information (RAI)

Additional Information M77222 and Nos. M77222 (TAC Nos. M77223)," letter and M77223)," letter October 7, dated October dated 7, 1994.

1994.

E.2-13 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 Power Station Applicant's Environmental Environmental Report Report {I' Operating License license Renewal Stage "-'

E.2-13 Liparulo, N.

N. J., (Westinghouse Corporation) to NRC Document (Westinghouse Electric Corporation) Document Control Desk, "Submittal of Material "Submittal Material Pertinent to the AP600 Design Certification Certification Review," letter dateddated December 15,1992.

15, 1992.

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

April 1995. .

E.2-15 U.S. Nuclear Regulatory Commission, NUREG-1560, NUREG-1 560, Individual Plant Examination Examination Program:

Program: Perspectives on Reactor Safety and Plant Performance, Performance, Volume 2, December December 1997. . .

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

E.2-17 Pilgrim Nuclear Power Station, Individual Plant Examination Examination (IPE)

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

(IPEEE)

Report, July 1994.

E.2-19 Commission, NUREG/BR-0184, U.S. Nuclear Regulatory Commission, NUREG/BR-01 84, Regulatory Analysis Technical Evaluation Handbook, January 1997.

QW E.2-14 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 Pilgrim Power Station Applicant's Applicant's Environmental Environmental Report Report Operating Operating License License Renewal Stage Stage Table Table E.2-1 E.2-1 Summary of Phase II SAMA Candidates Summary of Phase II SAMA Candidates Considered Considered in in Cost-Benefit Cost-Benefit Evaluation Evaluation Upper Phase Off-Slte Phase II II l Result Result ofof Potential Potential CDF >

Estimated Bound Estimated SAMA ID SAMA SAMA Enhancement Dose Conclusion SAMAID Enhancement* Reduction Benefit Estimated Cost Reduction Benefit Improvements Improvements Related Related to to Accident Accident Mitigation Mitigation Containment Containment Phenomena Phenomena 001 001 Install Install an an SAMA SAMAwouldwould decrease decrease 4.70%4.70% 4.60%

4.60% $43,639

$43,639 $261,832

$261,832. $5,800,000

$5,800,000 Not Not cost cost independent independent the probability of loss the probability of loss effective effective method method of of of of containment containment heat heat suppression suppression pool pool removal.

removal.

cooling.

cooling. I Basis Basisfor forConclusion:

Conclusion:

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

PNPS.

002 002 Install Install aa filtered filtered SAMA SAMAwouldwould provide provide 0.00%

0.00% 0.00%

0.00% $0

$0 $0

$0 $3,000,000

$3,000,000 Not Not cost cost containment containmentvent vent an alternate an alternatedecay decay effective effective to to provide provide fission fission heat heatremoval removal method method product product for non-ATWS for non-ATWS events, events, scrubbing.

scrubbing. with with fission fission product product Option Option 1:1: Gravel Gravel scrubbing.

scrubbing.

Bed Bed Filter Filter Option Option 2:2: Multiple Multiple Venturi Scrubber Venturi Scrubber -

Basis BasisforforConclusion:

Conclusion:

Successful Successfultorus torusventing venting accident accidentprogression progressionsource sourceterms terms are arereduced reduced bybyaafactor factorofof22toto reflect reflectthe the additional filtered capability. The cost of implementing this SAMA at Peach Bottom was estimated to be $3 additional filtered capability. The cost of implementing this SAMA at Peach Bottom was estimated to be $3 million. Therefore,this million. Therefore, this SAMA SAMAisisnot notcost costeffective effectiveforforPNPS.

PNPS.

E.2-15 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 Pilgrim Nuclear Nuclear Power Power Station Station Applicant's Environmental Applicant's Environmental Report Report Operating Operating License License Renewal Renewal Stage Stage Table Table E.2-1 E.2-1 Summary of Phase II SAMA Candidates Summary of Phase II SAMA Candidates Considered Considered in Cost-Benefit Cost-Benefit Evaluation Evaluation (Continued)

(Continued)

Upper Phase Off-5lte Phase II II Result Result ofof Potential Potential CDF CDF off-Site Estimated Estimated Upper BoundEsti Bound Estimated Estimated ated Conclusion SAMA ID SAMA Enhancement Reduction Dose dos Benefit Estimated Conclusion SAMAID Enhancement Reduction Benefit Estimated Cost Cost ReductionBnet Reduction Benefit 003 003 Install Install a a Assuming Assuming that that injection injection 0.50%

0.50% 1.19%

1.19% $10,283

$10,283 $61,701

$61,701 >$2,000,000

>$2,000,000 Not Not cost cost containment containment ventvent is available, is available, this this SAMA SAMA effective effective large large enough enough to to would would provide provide alternate alternate remove removeATWSATWS decay decay heat heat removal removal in in decay decay heat.

heat. an an ATWS ATWS event.

event.

Basis Basis for for

Conclusion:

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

million.

Therefore, Therefore, this this SAMA SAMA is is not not cost cost effective effective for for PNPS.

PNPS.

004 004 Create Create a a large large SAMA SAMA wouldwould ensure ensure 0.00%

0.00% 48.62%

48.62% $436,759

$436,759 $2,620,551

$2,620,551 >$100

>$100 million million NotNot cost cost concrete concrete crucible crucible that molten that molten corecore debris debris effective effective

- with with heat heat removal removal escaping escaping fromfrom thethe ,

potential potential under under vessel vessel would would be be the the base base mat mat to to contained contained within within thethe contain molten contain molten crucible.

crucible. The The water water core core debris.

debris. cooling cooling mechanism mechanism would would cool cool the the molten molten core, preventing core, preventing aa melt-through melt-through of ofthe the base mat.

base mat.

Basis Basis for for

Conclusion:

Containment Containment failurefailure due dueto to core-concrete core-concrete interactions interactions (not (not including including liner linerfailures) failures) was was eliminated eliminated to to conservatively assess the benefit of this SAMA.

conservatively assess the benefit of this SAMA. The cost The costof of implementing implementing this this SAMA SAMA at at ANO-2 ANO-2 waswas estimated estimated toto be be $100

$100 million.

million.

Therefore, Therefore, this this SAMA SAMA is is not not cost cost effective effective for for PNPS.

PNPS.

E.2-16 E.2-16

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

• .r-.

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR

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

Upper Phase IIII Off-5lte Phase SAMA Result of Result of Potential Potential CDF Estimated Bound Estimated SAMA ID SAMA Dose Conclusion SAMAID Enhancement Enhancement Reduction Benefit Estimated Cost

.i Reduction Benefit 005 005 Create a water-Create SAMA would contain 0.00% 48.62% $436,759 $2,620,551 $19,000,000 Not cost cooled rubble bed cooled rubble bed molten core debris effective on the pedestal.

on the pedestal. dropping on to the and would pedestal and 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.

benefitofthis SAMA. The cost of implementing implementing this SAMA at ANO-2 was estimated to be $19 million.

Therefore, this SAMA Therefore, SAMA is not cost effective for PNPS.

006 006 Provide SAMA SAMA would provide 0.00%

0.00% 0.07% $2,153 $12,915 >$1,000,000 Not cost modification for modification for intentional intentional flooding of . effective flooding the flooding the the upper drywell drywell head drywell head.

drywell head. such that if high drywell drywell temperatures temperatures occurred, occurred, the drywell drywell head seal seal would not fail.

Basis for

Conclusion:

Drywell head failures due to high temperature temperature were eliminated to conservatively assess the benefit of this SAMA.

SAMA. The cost of implementing implementing this SAMASAMA was estimated estimated to be greater greater than than $1

$1 million by engineering engineering judgment. Therefore, Therefore, this SAMA SAMA is not cost effective for PNPS.

E.2-17 E.2-17

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

.J J

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Pilgrim Nuclear Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 E.2-1 .

Summary of Phase II SAMA Summary SAMA Candidates Candidates Considered Considered in Cost-Benefit Cost-Benefit Evaluation (Continued)

(Continued)

Upper Off-5lte*

Phase IIII Phase Result of Potential CDF CDF Off-Site Estimated Estimated Upper Bound Estimated Estimated SAMA Dose Conclusion SAMAID SAMA ID Enhancement Enhancement Reduction Reduction Benefit Estimated Estimated Cost Reduction RedutionBenefit Benefit 007 Enhance fire SAMA would improve SAMA 0.00% 1.16%

1.16% $10,763 $64,577 >$2,500,000 Not cost protection system fission product effective and SGTS scrubbing in severe hardware 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. implementing this SAMA SAMA.. The cost of implementing SAMA was estimated to be greater than $2.5 million by engineering judgment.

judgment. Therefore, Therefore, this SAMA SAMA is not cost effective for PNPS.

E.2-18 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 E.2-1 Summary of Phase IIII SAMA Considered in Cost-Benefit SAM A Candidates Considered Cost*Benefit Evaluation (Continued)

(Continued) off-site Upper Upper Off-8lte Phase II Result of Potential CDF CDF Estimated Estimated Bound Estimated Estimated SAM A Dose Conclusion SAMA ID SAMAID Enhancement Enhancement Reduction Benefit Estimated

. Estimated Cost Conclusion Reduction ReductionBefi

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

and concrete attack attack would not occur.

Basis for

Conclusion:

Containment Containment failure due to core-concrete core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA.

SAMA. The cost of implementing implementing this SAMA SAMA was estimated to be greater greater than $5 million by engineering judgment. Therefore, this SAMA SAMA is not cost effective for PNPS.

E.2-19 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 Applicant's Environmental Report Operating License Renewal Stage Operating Table E.2-1 E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Summary (Continued)

Evaluation' (Continued)

Upper Off-5lte Phase IIII Result of Potential Potential CDF CDF OffEstimate Estimated Estmaed d Upper Bound Bondostmaed Estimated Estimated Conclusion SAMA Dose Conclusion SAMA SAMAIDID SAMA Enhancement Enhancement Reduction Reduce on Benefit Estimated Estimated Cost Cost Reduction RedutionBenefit Benefit 009 Install aa passive

,Install SAMA SAMA would decrease 5.05% 4.70% $44,037 $264,219 $5,800,000

$5,800,000. Not cost containmentspray containment spray the probability of loss effective system. of containment containment heat heat removal.

Basis for

Conclusion:

The CDF CDF contribution from loss of the drywell spray mode mode of RHR was eliminated to conservatively assess the benefit of this SAMA.

SAMA. The cost of implementing this SAMA SAMA at Quad Cities was estimated to be $5.8 million. Therefore, Therefore, this SAMA 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 containment over-over-containment. pressurization failure.

Conclusion:

Energetic containment failure modes (DCH, Basis for

Conclusion:

over-pressurization) were eliminated to (DCH, steam explosion, late over-pressurization) conservatively assess the benefit of this SAMA.

SAMA. The cost of implementing this SAMA SAMA at Quad Cities and at an ABWR ABWR was estimated estimated to be $12 million. Therefore, this SAMA SAMA is is not cost effective for PNPS.

E.2-20

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

~..

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

Candidates Considered in Cost-Benefit (Continued)

Upper Off-8lte Phase IIII Phase SAMA Result of Potential CDF lt Estimated Estimated Estmatd und Bound BunoEsimaed Estimated Estimated Conclusion Dose Conclusion SAMA SAMAID ID Enhancement Enhancement Reduction Reduction Benefit Benefit Estimated Estimated Cost Cost Reduction RedutionBenefit Benefit 011 Increase the Increase SAMA would prevent SAMA 0.00%

0.00% 0.43%

0.43% $4,305 $25,831

$25,831 >$5,000,000

>$5,000,000 Not Not cost cost depth of the mat melt-through.

base mat melt-through. effective effective concrete base mat or use an an alternative concrete material concrete material to ensure melt-melt-through does not occur. l Containment failure due to base mat Basis for

Conclusion:

Containment melt-through was eliminated to mat melt-through to conservatively assess thethe benefit benefit of this SAMA.

SAMA. The cost of implementing SAMA was estimated to be greater than $5 million implementing this SAMA engineering judgment. Therefore, million by engineering Therefore, this SAMA SAMA is not not cost effective for PNPS.

012 a reactor Provide a SAMA would provide SAMA 0.00%

0.00% 0.22% $3,229 $19,373 $2,500jO00

$2,500,000 Not cost Not vessel exterior the potential to cool aa effective effective cooling system. molten core before it causes vessel failure, if the lower head could be submerged submerged in in water.

Basis for

Conclusion:

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

E.2-21 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 Pilgrim Nuclear Nuclear Power Power Station Station Applicant's Applicant's Environmental Environmental Report Report Operating Operating License License Renewal Renewal Stage Stage Table Table E.2-1 E.2-1 Summary of Phase II SAMA Candidates Summary of Phase II SAMA Candidates Considered In Considered in Cost-Benefit Cost-Benefit Evaluation (Continued)

Evaluation'(Continued)

Upper Phase Off-5ite Phase 11 II Result Result of of Potential Potential CDF CDF Offite Estimated Est ma ed Estimated Bound Bo ndost Estimated Bound Estimated ma ed C onclusion SAMA ID SAMA M Enhancement Reduction Dose Conclusion SAMAID Enhancement Reduction Reduction Benefit Benefit Estimated Estimated Cost Cost Reduction Redu tionBenefit Benefit 013

,013 Construct Construct a a SAMA would SAMA would provide provide a a 0.00%

0.00% 1.16%

1.16% $10,763

$10,763 $64,577

'$64,577 >$2,000,000

>$2,000,000 Not Not cost cost building building method method to to effective effective connected connected to to depressurize depressurize primary primary containment containment and and containment containment that that reduce reduce fission fission product product is is maintained maintained atat aa release.

release.

vacuum.

vacuum.

Basis Basis for for

Conclusion:

FailureFailure ofof the the reactor reactor building building toto contain contain releases releases was was eliminated eliminated toto conservatively conservatively assess assess thethe benefit benefit ofof this this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom.

SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is Therefore, this SAMA is not not cost cost effective effective for for PNPS.

PNPS.

014 014 2.g.

2.g. Dedicated Dedicated SAMA would SAMA would decrease decrease 4.70%

4.70% 4.60%

4.60% $43,639

$43,639 $261,832

$261,832 $5,800,000

$5,800,000 Not Not cost cost Suppression Suppression PoolPool the probability the probability of of loss loss effective effective Cooling Cooling of containment of containment heat heat removal.

removal.

Basis Basis for for

Conclusion:

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

PNPS.

015 015 3.a.

3.a. Create Create a a SAMA SAMA increases increases time time 0.00%

0.00% 26.10%

26.10% $205,571

$205,571 $1,233,428

$1,233,428 $8,000,000

$8,000,000 Not Not cost cost larger larger volume volume in in before containment before containment effective effective containment.

containment. failure failure and and increases increases time for recovery.

time for recovery.

Basis Basis for for

Conclusion:

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

$8 million.

million.

Therefore, Therefore, this this SAMA SAMA is is not not cost cost effective effective for for PNPS.

PNPS.

E.2-22 E.2-22

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

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

Upper Off-5lte Phase II Result of Potential CDF CDF Off-Site Estimated Upper Bound EstimaEsE Estimated SAMA SAMA .-.. Dose Esimtdoond Esiatd Conclusion Conclusion SAMA ID SAMAID Enhancement Enhancement Reduction Benefit Benefit Estimated Estimated Cost:

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

capability (sufficient (sufficient pressure to withstand severe accidents).

accidents). ".

Basis for

Conclusion:

Energetic containment failure modes (DCH, (DCH, steam explosion, late over-pressurization) over-pressurization) were eliminated to conservatively assess the benefit of this SAMA.

SAMA. The cost of implementing this SAMA SAMA at Quad Cities and at an ABWR was estimated estimated to be $12 million. Therefore, this SAMASAMA is not cost effective for PNPS.

017 3.c. Install This SAMA SAMA addresses 0.00% 0.00%

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

Basis for ConclusIon:

Conclusion:

Vacuum breaker failures and suppression pool pool scrubbing failures were eliminated to conservatively assess the benefit of this SAMA.

SAMA. The cost of implementing implementing this SAMA SAMA at Peach BottomBottom was estimated estimated to be greater greater than $1 $1 million.

Therefore, this SAMA is not cost effective for PNPS.

Therefore.

E.2-23 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 Nuclear Power Power Station Applicant's Applicant's Environmental Environmental Report Report Operating Operating License license Renewal Renewal Stage Stage

_ Table E E.2-1 . 2 - 1 . . .

Summary of Summary of Phase II SAMA Candidates Considered SAMA Candidates Cost-Benefit Evaluation Considered in Cost-Benefit Evaluation (Continued)

(Continued)

Upper Off-5lteOff- itesti m ted Upper Phase II-Phase 11 Result of Potential CDF CDF OffSite Estimated Estimated Und

. Bound Estimated Estimated SAMA Dose Conclusion SAMA SAMAIDID Enhancement Enhancement Reduction Reduction Rducion Benefit Esim Estimated d Cost Cost Reduction Redu tionBenefit Benefit 018 3.d. Increase the This SAMA SAMA would 0.00%

0.00% 0.07% $2,153 $12,915 $12,000,000 Not cost cost temperature temperature reduce the potential for effective margin for seals.

margin containment failure containment under adverse under conditions.

temperature drywell seal failure was eliminated Containment failure due to high temperature Basis for

Conclusion:

Containment conservatively assess the eliminated to conservatively benefit of this SAMA.

SAMA. The cost of implementing SAMA at Quad Cities and at an ABWR were implementing this SAMA were estimated to be $12 million and and was attainable benefit, even without a detailed cost estimate. Therefore, this SAMA judged to exceed the attainable SAMA is not cost C<:?st effective for PNPS.

019 5.b/c. Install a SAMA would provide SAMA 0.00%

0.00% 0.00%

0.00% $0 $0 $3,000,000 Not cost filtered vent an alternate decay effective heat removal method non-ATWS events, for non-ATWS with fission product 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 implementing this SAMA at Peach Bottom Bottom was estimated to be $3

$3 million. Therefore, this SAMA is 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 Power Station Applicant's Applicant's Environmental Environmental Report Report Operating License Renewal Renewal Stage Stage Table Table E.2-1 E.2-1 Summary Summary of of Phase Phase IIII SAMA SAMA Candidates Candidates Considered Considered in in Cost-Benefit Evaluation (Continued)

Cost-Benefit Evaluation (Continued)

Upper Phase II Result Off-5ite Phase II Result of of Potential Potential CDF CDF i Estimate Estmaed Estimatedd BuEstimated Bondostmaed Bound Estimated

Conclusion:

SAMA ID SAMA Enhancement Reduction Dose De Benefit Estimated Cost Conclusion SAMAID Enhancement Reduction Reduction Benefit Estimated Benefit Cost Reduction Benefit 020 020 7.a.

7.a. Provide Provide aa SAMA SAMA would would provide provide 0.00%

0.00% 0.07%

0.07% $2,153

$2,153 $12,915

$12,915 >$1,000,000

>$1,000,000 Not Not cost cost method method of of drywell drywell intentional intentional flooding flooding of of effective effective head flooding.

h-ead flooding. the the upper upper drywell drywell headhead such such that that if if high high drywell drywell temperatures temperatures occurred, occurred, the the drywell drywell head head sealseal would not would not fail.

fail.

Basis Basis forfor

Conclusion:

Drywell head failures Drywell head failures due due to to high high temperature temperature werewere eliminated eliminated toto conservatively conservatively assess assess the the benefit benefit of ofthis this SAMA.

SAMA. The The cost cost of of implementing implementing this this SAMA SAMA waswas estimated estimated to to be be greater greaterthan than $1

$1 million million by by engineering engineering judgment.

judgment. Therefore, Therefore, this this SAMA SAMA is is not not cost cost effective effective forfor PNPS.

PNPS.

021 021 13.a.

13.a. Use Use This This SAMA SAMA provides provides 0.00%

0.00% 1.16%

1.16% $10,763

$10,763 $64,577

$64,577 >$2,500,000

>$2,500,000 Not Not cost cost alternate alternate method method the capability the capability to to use use effective effective of ofreactor reactor building building firewater firewater sprays sprays in in the the spray.

spray. reactor reactor building building to to mitigate mitigate release release of of fission fission products products intointo the reactor building the reactor building following following anan accident.

accident Basis Basis for for

Conclusion:

Failure Failure ofofthe the reactor reactor building building to to contain contain releases releases was was eliminated eliminated toto conservatively conservatively assess assess the the benefit benefit of ofthis this SAMA. The cost of implementing SAMA. The cost of implementing this this SAMA SAMAwas was estimated estimated to to be be greater greater than than $2.5

$2.5 million million by by engineering engineeringjudgment.

judgment. Therefore, Therefore, this this SAMA SAMA is is not notcost cost effective effective forfor PNPS.

PNPS.

E.2-25 E.2-2S

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

Cost-Benefit Evaluation (Continued)

Upper Off-Slte Phase II II Result of Potential CDF CDF OffSte Estimated Estimated Est ma ed Bound Bound Bo ndost Estimated Estimated ma ed C onclusion Conclusion SAMA Dose SAMA SAMAIDID SAMA Enhancement Enhancement Reduction Reduction Rductios Benefit Benefit Estimated Estimated Cost Cost ReductionRedu tionBenefit Benefit 022 14.a. Provide a SAMA SAMA would allow the 0.00% 22.48%

22.48% $204,495

$204,495* $1,226,971 $2,500,000 - Not cost means ofof flooding flooding debris to debris to be be cooled.

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.

SAMA. The cost of implementing implementing this SAMASAMA at Quad Cities was estimated to be $2.5 million. Therefore, this SAMA SAMA is 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 reactor cavity debris coolability, debris coolability, effective flooding system. reduce core concrete interaction, and interaction, and provide fission product scrubbing.

Basis for

Conclusion:

Containment 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 SAMA at ANO-2 implementing this SAMA ANO':2 was estimated to be $8.75 million.

Therefore, Therefore, this SAMA SAMA is is not cost effective for PNPS.

024 Add ribbing to the This SAMA SAMA would 0.00% 26.10%

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

E.2-26 E.2-26

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

(Continued)

Upper Off-5ite Off-Site Estimated Phase II Result of Potential CDF Estimated Estmaed Upper Bound Bondostmaed Estimated Estimated Conclusion SAMA Dose Conclusion SAMA ID SAMAID SAMA Enhancement Enhancement Reduction Benefit Estimated Estimated Cost Cost Reduction Benefit Improvements Improvemen~ Related to to Enhanced Enhanced ACIDC AC/DC Reliability/Availability R 025 025 Provide additional SAMA would SAMA would ensure ensure 1.39%

1.39% 2.79%

2.79% $24,393 $146,356

$146,356 $500,000

$500,000 Not cost cost DC battery DC battery longer battery longer battery effective effective capacity.

capacity. capability capability during an during an SBO, which would extend HPCIIRCIC extend HPCW/RCIC operability and allow allow more time more time for for AC AC power recovery.

Basis for

Conclusion:

The BaSis The time available to recover offsite power before before HPCI RCIC are lost was changed from 14 HPCI and RCIC 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> to 24 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during SBO hours SBO scenarios toto conservatively assess assess the benefit of this SAMA.

SAMA. The cost of implementing implementing this SAMA SAMA was estimated to be $500,000 be $500,000 by by engineering engineering judgment. Therefore, Therefore, this SAMA SAMA is is not cost effective for PNPS.

026 026 Use fuel Use fuel cells cells SAMA would extend SAMA 1.39%

1.39% 2.79%

2.79% $24,393 $146,356 >$2,000,000 Not cost instead instead of of lead-lead- DC DC power power availability availabinty in in effective acid acid batteries.

batteries. an SBO, an SBO, which which would HPCI/RCIC extend HPCIIRCIC operability and allow more time for ACAC power recovery.

Basis for

Conclusion:

The time available to recover offsite power before HPCI and RCIC 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 during SBO scenarios scenarios to conservatively conservatively assess the benefit of this SAMA. SAMA. The cost of implementing implementing this SAMA SAMA at Peach BottomBottom was was estimated estimated to be greater than $2 $2million. Therefore, Therefore, this SAMA SAMA is not cost effective for PNPS.

E.2-27 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 Pilgrim PowerStation NuclearPower Station Applicant's Environmental Report Applicant's Environmental Report Operating License Operating Renewal Stage License Renewal Stage Table*E.2-1 Table E.2-1 Summary of Phase II SAMA Candidates Considered in Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaiuation (Continued)

Cost-Benefit Evaluation (Continued)

Upper Off-5lte Bound Estimated Phase 11 Phase II SAMA Result of Result of Potential Potential CDF CDF OffSite .Estimated Estimated Esim tdoo BoundndEsi Estimated atd Conclusion SAMA Enhancement Reduction Dose Rduction Benefit Estimated Cost Conclusion SAMA SAMAID ID Enhancement Reduction Benefit Estimated Cost Reduction Redu tionBenefit Benefit 027 027 Modification for Modification for SAMA would increase SAMA would increase 4.65%

4.65% 1.91%

1.91% $19,761

$19,761 $118,568

$118,568 $500,000

$500,000 Not cost Not cost Improving Improving DCDC Bus Bus reliability of AC power reliability of AC power effective effective Reliability Reliability injection capability.

and injection and capability.

Basis for Basis

Conclusion:

The for

Conclusion:

The CDF contribution due CDF contribution of DC loss of to loss due to buses D16 DC buses D16and eliminated to was eliminated D17 was and D07 assess the to assess ofthis benefit of the benefit this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment. Therefore, this SAMA is not SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment. Therefore, this SAMAis not cost cost effective for PNPS.

effective for PNPS.

028 Provide 16-2.1. Provide 2.i. 16- SAMA includes SAMA includes 1.39%

1.39% 2.79%

2.79% $24,393

$24,393 $146,356

$146,356 $500,000

$500,000 Not Not cost cost hour SBO hourSBO capability to improved capability improved to effective effective injection.

injection. cope with longer SBO cope with longer SBO l scenarios.

scenarios. l Basis for Basis

Conclusion:

The for

Conclusion:

The time available to time available to recover power before offsite power recover offsite and RCIC HPCI and before HPCI are lost RCIC are lost was changed from was changed from 14 hours to 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> to 24 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> hours during scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to SBO scenarios during SBO to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to

$500,000 by be $500,000 be by engineering Therefore, this judgment. Therefore, engineering judgment. SAMA is this SAMA not cost is not for PNPS.

effective for cost effective PNPS.

E.2-28 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 E.2-1 Table E.2-1 Summary of Phase"Phase II SAMA SAMA Candidates Considered Considered in In Cost-Benefit (Continued)

Cost-Benefit Evaluation (Continued)

Upper Off-5lte Phase II Result of Potential Result CDF OffEstimate Estmaedd Estimated Bound Bondostmaed Estimated Estimated Conclusion SAM A Dose Conclusion SAMA ID SAMAID SAM Enhancement

• • Enhancement Reduction Benefit Estimated Estimated Cost ReductionBefi Reduction Benefit 029 9.b. Provide an SAMA would This SAMA 2.22% 5.06% $44,281 $265,687 >$2,000,000 Not cost alternate pump provide asman, a small, effective power source. power dedicated power source such as aa dedicated diesel or gas turbine for the feedwater or condensate pumps so so that they do not rely on offsite power.

Basis for

Conclusion:

The CDF CDF contribution due to failure of the SBO diesel was eliminated to conservatively assess the benefit of of this SAMA.

SAMA. The cost of implementing implementing this SAMA SAMA at Peach Bottom Bottom was estimated to be greater than $2 million. Therefore, Therefore, this SAMA SAMA is not cost effective for PNPS.

030 9.g.

9;g. Enhance SAMA SAMA would provide 11.10%

11.10% 8.47%

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

Basis for

Conclusion:

The CDF CDF contribution due due to to loss of MCCs B17, B18, MCCs B17, 818, and B15 815 was was eliminated eliminated to conservatively assess the benefit of this SAMA.

SAMA. The cost of implementing implementing this SAMA SAMA was estimated to be $146,120 by engineering 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 Power Station Applicant's Environmental Environmental Report Operating License Renewal Stage Stage Table E.2-1 Summary of Phase IIII SAMA Summary SAMA Candidates Candidates Considered in Cost-Benefit Cost-Benefit Evaluation (Continued)

(Continued)

Upper 11 Phase II SAMA Result of of Potential Potential CDF Off-SiteUpe Off-5ite ose Estimated Bound Estimated CDF Estimated Bound Estimated SAMA Dose Conclusion SAMA ID SAMAID . Enhancement Enhancement Reduction ReductionBeet Benefit Estimated Estimated Cost Reduction Benefit 031 10.a. Add aa 10.a. This SAMA SAMA addresses 24.3%

24.3% 16.16%

16.16% $150,504 $903,025 $3,000,000

$3,000,000 Not cost dedicated DC the use of a diverse DC effective power supply. power system such power such as as an additional battery or fuel cell for the purpose of providing motive power to certain components (e.g.,

RCIC).

Basis for for

Conclusion:

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

032 10.b. Install This SAMA SAMA addresses 24.3%

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 as divisions. an additional battery or or fuel cell for the purpose of providing motive power to certain components (e.g.,

(e.g.,

RCIC).

RCIC).

Basis for

Conclusion:

The CDF CDF contribution due to loss of DC DC Bus 'B'

'B' was eliminated eliminated to conservatively assess the benefit of this SAMA. The cost of implementing implementing this SAMA SAMA at Quad Cities was estimated to be $3

$3 million. Therefore, Therefore, this SAMA SAMA is is not cost effective for PNPS.

forPNPS.

E.2-30

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

(Continued)

Upper Off-8ite Phase II Phase Result of Potential CDF CDF Off-Site Estimated Estimated Bound Estimated Estimated SAMA Dose Conclusion SAMA SAMAIDID Enhancement Enhancement Reduction Dose Benefit Estimated Estimated Cost Cost., Conclusion ReductionBeet Reduction Benefit Benefit 033 10.c. Install fuel SAMA would extend SAMA 1.39% 2.79% $24,393 $146,356 >$2,000,000

>$2,000,000 Not Not cost .

cells.

cens. DC power availability in effective an SBO, SBO, which would extend HPCI/RCIC HPCIIRCIC operability and allow more more time for AC power 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.SAMA. The cost of implementing SAMA at Peach Bottom implementing this SAMA was estimated to be greater than $2

$2 million. Therefore, this SAMA SAMA is not cost effective for PNPS.

034 10.d. Enhance This SAMA SAMA would 4.65% 1.91%

1.91% $19,761 $118,568 $13,000 Retain procedures to improve DCDC power power make use of DC DC availability.

bus cross-ties.

cross-ties .

Basis for

Conclusion:

The CDFCDF contribution due to loss of DC buses 016 D16 and D17 017 was eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA 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 Environmental Report Report Operating License Renewal Stage Table E.2-1 Summary of Phase IIII SAMA SAMA Candidates Considered Considered in Cost-Benefit Cost-Benefit Evaluation (Continued)

(Continued)

Upper Phas IIResut o Potntil CD Of~itUpper Off-8ite Phase II Phase SAMA Result of Potential CDF Dose Estimated Estimated Bound Estimated Estimated Conclusion SAMA Dose Conclusion SAMA SAMAID ID Enhancement Enhancement Reduction Reduction Reduction Benefit Estimated Estimated Cost Reduction RedutionBenefit Benefit 035 10.e. Extended Extended SAMA would SAMA would extend extend 1.39%

1.39% 2.79% $24,393

$24,393 $146,356 $500,000

$500,000 Not cost SBO provisions. DC power availability in effective an SBO, which would HPCI/RCIC extend HPCIIRCIC operability and allow more time for AC power recovery.

Basis for

Conclusion:

The time available to recover offsite power before HPCI 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 HPCI and RCIC hours during SBO scenarios to conservatively assess the benefit of this SAMA.SAMA. The cost of implementing implementing this SAMA SAMA was estimated to

$500,000 by engineering judgment. Therefore, this SAMA is not cost effective for PNPS.

be $500,000 Improvements Improvements in Identifying and Mitigating Containment Bypass 036 Locate RHR SAMA would SAMA would prevent prevent 0.33% 0.21%

0.21% $2,749 $16,497 >$500,000

>$500,000 Not cost inside ISLOCA ISLOCA outside effective containment. containment. .

Basis for

Conclusion:

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

037 Increase Increase SAMA SAMA could reduce reduce 0.54%

0.54% 0.38%

0.38% $4,025 $24,148 $100,000

$100,000 Not cost frequency of valve ISLOCA frequency.

ISLOCA effective leak testing. _ _

Basis for

Conclusion:

The CDF CDF contribution due to ISLOCA ISLOCA was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA engineering judgment. Therefore, SAMA was estimated to be $100,000 by engineering Therefore, this SAMA 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 Power Station Applicant's Environmental Environmental Report License Renewal Stage Operating license Stage E.2-1 Table E.2-1 Summary of Phase Summary Phase IIII SAMA Candidates Candidates Considered Considered In Cost-Benefit Evaluation Evaluation (Continued)

(Continued)

Upper Off-Slte II Phase II SAMA Result of Potential CDF CDF OffSit Estimated Estimated Estmaed Upper Bound Bondostmaed Estimated Estimated Conclusion Dose Conclusion SAMA ID SAMAID Enhancement Enhancement Reduction Dose ReductionBeet Benefit Estimated Estimated Cost Cost Reduction Benefit 038 8.e. Improve This SAMA SAMA would 0.00% 0.00%

0.00% $0

$0 $0

$0 >$2,000,000 Not cost Not cost MSIV MSIV design. decrease the decrease the likelihood likelihood effective effective of containment bypass scenarios.

Basis for Basis for

Conclusion:

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

Improvements Improvements Related to Core Cooling System 039 Install an SAMA SAMA would allow 0.00% 0.00% $0

$0 $0

$0 $135,000 Not cost independent independent continued inventory in in effective diesel for the CST CST during an SBO.SBC.

makeup pumps.

Basis for

Conclusion:

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

Therefore, this SAMA SAMA isis not cost effective for PNPS.

E.2-33

• J 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 Pilgrim Power Station Nuclear Power Station Applicant's Environmental Applicant's Environmental Report Report License Renewal Operating License Operating Renewal Stage Stage Table Table E.2-1 E.2-1 Summary of Phase 11 Summary II SAMA Candidates Considered SAMA Candidates Considered in Cost-Benefit Cost-Benefit Evaluation Evaluation (Continued)

(Continued)

Upper Off-5lte Phase Phase II II . Result of Potential CDF CDF Offt Estimated Estimated Estmaed Upped Bound Bondostmaed Estimated Estimated Conclusion SAMA Dose Conclusion SAMAID SAMA ID S Enhancement Enhancement Reduction Reduction De Benefit Benefit Estimated Estimated Cost Cost Reduction ReductionBefi Benefit 040 Provide an SAMA would reduce SAMA 3.15%

3.15% 1.97%

1.97% $18,369 $110,212 >$2,000,000 Not cost additional high frequency of core melt melt effective pressure injection from small LOCA andand pump with SBO SBO sequences.

sequences.

independent independent diesel.

Basis for

Conclusion:

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

041 Install SAMA would allow SAMA 3.15% 1.97%

1.97% $18,369

$18,369 $110,212 >$2,000,000 Not cost cost independent AC makeup capabilities effective high pressure during transients, small small injection system. LOCAs, and SBOs.

SBOs.

contribution due to Basis for

Conclusion:

The CDF contribution to failure failure ofthe of the HPCI HPCI system waswas eliminated to to conservatively assess the benefit of conservatively assess this SAMA. The cost of implementing Bottom was estimated to be greater SAMA at Peach Bottom implementing this SAMA million. Therefore, greater than $2 million. this SAMA 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 Nuclear Power Station Environmental Report Applicant's Environmental Operating License Renewal Stage Table E.2-1 E.2-1 Summary Summary of Phase"Phase II SAMA SAMA Candidates Considered in In Cost-Benefit Cost-Benefit Evaluation (Continued)

(Continued)

Upper Off-5iteOff-SiteUpper Phase II11 Phase SAMA SAMA Result of Potential CDF CDF off-Site Dose Estimated Estimated Bound Estimated Estimated

Conclusion:

Conclusion*

SAMA ID SAMAID Enhancement Enhancement Reduction De Benefit' . Estimated Benefit Estimated Cost Cost Reduction ReductionBeet Benefit 042 2.a. Install aa SAMA would improve SAMA improve 3.15% 1.97%

1.97% $18,369 $110,212 >$2,000,000 Not cost cost high passive high prevention of core prevention core melt melt effective effective pressure system. sequences by by providing additional high pressure capability to remove decay heat through an an isolation condenser type system.

Basis for

Conclusion:

The CDF contribution due to failure of the HPCI system was eliminated eliminated to conservatively assess the benefit of this SAMA.

SAMA. The cost of implementing implementing this SAMA SAMA was estimated to be greater than $2 million at Peach Bottom. Bottom. Therefore, Therefore, this SAMA SAMA is not cost effective for PNPS is PNPS...

043 2.d. Improved SAMA will improve 2.11%

2.11% 1.43%

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

Basis for

Conclusion:

The CDF 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 implementing this SAMASAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, Therefore, this SAMA SAMA is not cost effective for PNPS.

E.2-35

a 3 NRC - Applicant's Environmental Report SAMA Analysis

.J Exhibit No. NRC000001 J

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR Pilgrim Nuclear Power Station Applicant's Environmental Environmental Report Operating License Renewal Stage

. E.2-1 Table E.2-1 Summary Summary of Phase Phase IIII SAMA Candidates Candidates Considered Considered In Cost-Benefit Cost-Benefit Evaluation (Continued)

(Continued)

.. Upper Off-5lte Phase II11 Phase Result of Result of Potential Potential CDF CDF OffSt Estimated Estimated Bound Estimated Estimated SAMA SAMA Dose Esimtdoond Esiatd Conclusion SAMA ID SAMAID Enhancement Enhancement Reduction ReductionBeet Benefit Estimated Estimated Cost Reduction Benefit 044 2.e. Install an SAMA will SAMA will improve 3.15%

3.15% 1.97%

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 an additional system.

Basis for Basis for

Conclusion:

The CDF contribution due CDF contribution due to failure failure of of the HPCI HPCI system system was was eliminated eliminated to to conservatively conservatively assess the benefit of of this SAMA.

SAMA. The cost of implementing implementing this SAMA SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, Therefore, this SAMA SAMA is not cost effective for PNPS.

is 045 8.c. Add a a diverse SAMA will improve SAMA 3.15%

3.15% 1.97%

1.97% $18,369 $110,212 >$2,000,000 Not cost injection system. prevention of prevention of core core melt melt effective sequences by by providing additional additional.

injection capabilities.

Basis for

Conclusion:

Basis for

Conclusion:

The The CDF CDF contribution due to failure of the HPCIHPCI system was eliminated to conservatively assess the benefit of this SAMA.

SAMA. The cost of implementing implementing this SAMA at Peach Bottom Bottom was estimated to be greatergreater than $2 million. Therefore, Therefore, this SAMA SAMA is not cost effective for PNPS.

is .'

E.2-36

Exhibit No. NRC000001 It--- SAMA Analysis c

NRC - Applicant's Environmental Report 1

Pilgrim LR Proceeding 50-293-LR, 06-848-02-LR It-,

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

(Continued).

Phase IIII Phase 1

I Result of Potential CDF CDF OffSiteUpper Off-Site OffSite Estimated Estimated Upper Bound Bound Estimated Estimated C SAMA Dose Conclusion SAMA ID SAMA Enhancement Dose Benefit Estimated Cost SAMAID "

I.

Enhancement Reduction Reduction

~~~~~Reduction Estimated Bnft Benefit ______

Cost*

Improvements Improvements Related to ATWSATWS Mitigation 046 046 Increase SRV Increase SRV SAMA SAMA addresses the 1.51%

1.51% 0.92%

0.92% $10,600 $63,599

$63,599 $2,000,000 Not cost Not reseat reliability. risk associated with effective effective dilution of boron caused by the failure of the SRVs to reseat after SLC injection.

injection.

Basis for

Conclusion:

The CDF CDF contribution due to stuck open relief valves was eliminated to conservatively assess the benefit benefit of this SAMA.

this SAMA. The cost of implementing implementing this SAMA SAMA was estimated estimated to to be $2

$2 million million at at Peach Peach Bottom.

Bottom. Therefore, Therefore, this SAMASAMA is is not cost cost effective for PNPS.

047 11.a. Install an 11.a. SAMA would This SAMA 0.50%

0.50% 1.19%

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

E.2-37

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

(Continued)

Upper Phase II11 Result Off-5lte of Potential Potential CDF Offit Estimated SAMA SAMA Result of CDF Dose Esim tdoo Bound Estimated ndEsi Estimated atd Conclusion SAMA SAMAID ID Enhancement Enhancement Reduction Reduction Dos Benefit Benefit Estimated Estimated Cost ReductionBefi Reduction Benefit 048 048 Diversify Diversify An alternate means of An of 0.00%

0.00% 0.02% $2,153 $12,915 >$200,000 Not cost explosive valve explosive valve opening aa pathway opening pathway toto effective operation.

operation. the RPV RPV forfor SLC SLC system injection would improve the improve success the success probability for reactor reactor shutdown.

shutdown.

Basis for Basis for

Conclusion:

Common Common cause failure of SLC explosive valves was eliminated to conservatively assess the benefit of this SAMA. The cost of SAMA. of implementing implementing thisthis SAMA SAMA was estimated estimated to be greater than $200,000 by engineering judgment.

judgment. Therefore, this SAMA SAMA is is not not cost cost effective effective for for PNPS.

PNPS.

Other Improvements 049 Increase the Increase the SAMA reduces the SAMA 0.73%

0.73% 0.60% $5,300 $31,799 >$1,500,000

>$1,500,000 Not cost of SRVs reliability of SRVs consequences of consequences effective by adding signals medium break medium break LOCAs.

LOCAs.

to open them automatically.

Basis for

Conclusion:

The The CDF CDF contribution from SRVs SRVs failing to open in in medium medium LOCA LOCA sequences was eliminated to conservatively assess the benefit of this SAMA. SAMA. The cost of implementing implementing this SAMA SAMA was estimated to be greater than $1.5$1.5 million by by engineering engineering judgment. Therefore, Therefore, this SAMA 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 Power Station Environmental Report Applicant's Environmental License Renewal Stage Operating license Stage Table E.2-1 Table E.2-1 Summary of Summary Phase IIII SAMA Candidates of Phase Considered in Cost-Benefit Evaluation Candidates Considered Evaluation (Continued)

(Continued)

Upper Off-5lte Phase IIII Phase Result of Potential CDF CDF Offsit Dose Estimate Estimated d Upper Bound Estimated Estimated SAMA Conclusion SAMA ID SAMAID SAMA Enhancement Enhancement Reduction Dose Reduction:Bnei Reduction Benefit Estimated Estimated Costnclus Cost Benefit 050 8.e. Improve SRV SAMA would This SAMA 4.81%

4.81% 3.51%

3.51% $32,396 $194,378 >$2,000,000 Not cost design. improve improve SRV reliability effective thus increasing the likelihood that sequences could be mitigated using low-mitigated pressure heat removal.

Basis for

Conclusion:

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

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

water system.

Basis for

Conclusion:

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

implementing this SAMA Therefore, this SAMA SAMA is is not cost effective for PNPS E.2-39 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 Environmental Report Operating License Renewal Stage E.2-1 Table E.2-1 Summary of Phase Summary Phase IIII SAMA Candidates Candidates Considered Considered in In Cost-Benefit Evaluation (Continued)

(Continued)

Upper

'Off-5lte Phase IIII Phase SAMA Result of

' Result of Potential CDF CDF OffSite Estimated Estimated Upper Bound Estimated Estimated SAMA Dose Dose Etmtd B u dE i aed Conclusion SAMA ID SAMAID Enhancement Enhancement Reduction Reduction Reducton Benefit Bem!fit Estimated Estimated Cost Reduction Redu tionBenefit Benefit 052 052 Provide digital Provide digital Upgrade plant Upgrade plant 0.07%

0.07% 0.01%

0.01% $2,352 $14,109 >$100,000 Not cost large break large break LOCA LOCA instrumentation and instrumentation effective protection. logic to improve the logic capability to identify symptoms/precursors symptoms/precursors of a large break LOCA ofa LOCA (a break).

(a leak before break).

Basis for for

Conclusion:

The CDF CDF contribution due to large break LOCA LOCA was eliminated to conservatively assess the benefit of this SAMA.

SAMA. The cost of implementing implementing this SAMA SAMA was estimated to be greater greater than $100,000 by engineering judgment. Therefore, Therefore, this SAMA SAMA is is not cost effective for PNPS.

E.2-40 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 Environmental Report Applicant's Environmental license Renewal Stage Operating License Stage Table E.2-1 E.2-1 Summary Summary of Phase II !)AMASAMA Candidates Considered in Cost-Benefit Evaluation (Continued) (Continued)

Upper Phas IIRestto Potntil CF iteUpper

~ Off-~me PhaseS Phase II A ResuMA-of Resutt-of Potential CDF CDF Estimated Estimated Bound Estimated o SAMA Dose Conclusion SAMA SAMAIDID Enhancement Enhancement Reduction Dose Benefit Estimated Estimated Cost.

Cost. Conclusion ReductionBefi Reduction Benefit Improvements Improvements Related to IPE, IPE Update& IPEEE Insights Update& IPEEE 053 Control Control SAMA would This SAMA 3.61%

3.61% 2.24% $22,873 $137,237

$137,237 $300,000

$300,000 Not cost containment establish aa narrow effective venting within a pressure control band narrow band of to prevent rapid pressure containment containment depressurization depressurization when venting is implemented implemented thus avoiding adverse impact on the low pressure ECCS injection systems taking suction from the torus.

torus.

famng to recognize the need to vent the torus was reduced by a factor of 3 to operator failing Basis for

Conclusion:

The probability of the operator implementing this SAMA SAMA on CDF. The cost of implementing conservatively assess the benefit of this SAMA $300,000 by SAMA was estimated to be $300,000 by Therefore, this SAMA is engineering judgment. Therefore, Is not cost effective for PNPS.

E.2-41 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 Applicant's Environmental Environmental Report RePort Operating Operating License Renewal Stage Table E.2-1 Table E.2-1 Summary of Phase Summary Phase IIII SAMA Candidates Candidates Considered Considered in Cost-Benefit Evaluation (Continued)(Continued)

Upper Off-Slte Phase IIII Phase Result of Potential Potential CDF CDF Estimated Estimated Upper Bound Estimated Estimated SAMA Dose Conclusion SAMA ID SAMAID SAMA Enhancement Enhancement Reduction Reduction Dose Benefit Estimated Estimated Cost onclus on Reduction ReductionBeet Benefit 054 Install a bypass This SAMA SAMA would 0.28%

0.28% 0.33%

0.33% $3,627 $21,761 $1,000,000 Not cost to bypass switch to. reduce the core effective the low reactor damage frequency frequency pressure contribution from the interlocks of LPCI transients with stuck stuck or core spray open SRVs or LOCAs injection valves cases. Core Spray Spray and LPCI injection valves require aa low permissive 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 SAMA on CDF.

CDF. The cost of implementing this SAMASAMA at Dresden Dresden was estimated estimated to be $1

$1 million. Therefore, Therefore, this SAMA SAMA is not cost effective for PNPS. .'

is 055 Increase the SAMA would This SAMA 4.37% 6.63%

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

Basis for

Conclusion:

The CDF contribution from sequences involving common common cause failures of SSW and RBCCWRBCCW was eliminated to conservatively assess the benefit of this SAMA.

SAMA. The cost of implementing implementing this SAMA SAMA was estimated to be greater than $5 $5million by by engineering judgment. Therefore, this SAMA is 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 Pilgrim Nuclear Power Station Applicant's Environmental Environmental Report Operating License Renewal Stage Table E.2-1 Table E.2-1 Summary Summary of Phase" Phase II SAMA Candidates Candidates Considered Cost-Benefit Evaluation (Continued)

Considered In Cost-Benefit (Continued)

Upper Off-5lte II Phase II SAMA SAMA Result of Potential CDF Off-Site Estimated Estimated Upper Bound Estimated Estimated Dose Conclusion SAMA ID SAMAID S Enhancement Enhancement Reduction Reduction Reduction Benefit Estimated Estimated Cost Cost RedutionBenefit Benefit 056 Provide redundant This SAMA SAMA would 8.81%

8.81% 3.51%

3.51% $36,773 $220,639 $112,400

$112,400 Retain DC power DC improve reliability of supplies to DTV DTV the DTV valves and valves. enhance containment removal heat removal capability.

Basis for

Conclusion:

The CDF contribution contribution from sequences involving DC DC power supply failures to the DTV DTV valves was eliminated to conservatively assess the benefit of this SAMA.

SAMA. The cost of implementing implementing this SAMA SAMA was estimated to be $112,400 by engineering judgment.

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

poWer.

Basis for

Conclusion:

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

$26,000 by engineering judgment.

judgment.

. E.2-43 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 Environmental Report Operating License Renewal Stage Table E.2-1 E.2-1 Summary Summary of Phase II SAMA SAMA Candidates Considered In in Cost-Benefit Cost-Benefit Evaluation (Continued)

(Continued)

Upper Upper Off-~me Phase Phase 11 II Result of Potential Potential CDF CDF f-eeneit Estimated Buppenr Bound Cst Estimated SAMA Dose Conclusion SAMA SAMAIDID SAMA Enhancement Enhancement Reduction Reduction Dose Benefit Benefit Estimated Estimated Cost Cost Cnlso Reduction Benefit 058 Proceduralize the This SAMA would 4.92% 3.14%

3.14% $31,799 $190,797 $50,000 Retain operator action to provide the direction to feed B BI1 loads via restore B15 and B17 B17 B3 When WhenA5 A5 is is loads upon loss of A5 unavailable post- initiating events as long trip. Similarly, as A3 is available.

feed B2 loads via feed Additionally, it would Additionally, would B4 when A6 is B4 is provide the direction to to unavailable post restore B14 and B18 B18 trip.,

trip~ loads upon upon loss loss of A6 ..

initiating events as long as A4 is is available.

Basis for Basis for

Conclusion:

The CDF CDF contribution from sequences involving loss of 4160VAC 4160VAC safeguard bus A5 AS was conservatively eliminated eliminated to to assess the the benefit of this SAMA.

SAMA. The cost of implementing implementing this SAMA SAMA was estimated to be $50,000 by engineering engineering judgment.

059 Provide redundant This SAMA SAMA would 8.77%

8.77% 17.19%

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 BB Aand firewater cross-tie to cross-tie. LPCI loops A and B LPClloops B for reactor vessel injection and drywell spray.

Basis for

Conclusion:

Basis

Conclusion:

The CDF CDF contribution from sequences involving firewater injection failures was conservatively eliminated eliminated to assess the assess the benefit benefit of this this SAMA.

SAMA. The cost of implementing implementing this SAMASAMA was estimated to be $1,956,000 by engineering judgment.

Therefore, this Therefore, this SAMA SAMA is isnot cost cost effective for for PNPS 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 Pilgrim Nuclear Nuclear Power Power Station Station Applicant's Applicant's Environmental Environmental Report Report Operating Operating License License Renewal Renewal Stage Stage Table Table E.2-2 E.2-2 Sensitivity Sensitivity Analysis Analysis Results Results Upper Upper Upper Upper Upper Upper PhEstimated Estimated Bound Bound Estimaited Estimated Bound Bound Estimated Estimated Bound Bound Phase II Benefit Estimated Benefit Estimated Benefit Benefit Estimated Estimated Benefit Benefit Estimated Estimated II Estimated SAMASAMA SAMA Benefit*

Benefit Cost Benefit Benefit Benefit SAM A ----.-~

Cost Benefit 10 IDBase Line Base Lne Sensitivity Sensitivity Sensitivity BaseLine Base Line Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Case Case I1 Case Case I1 Case Case 22 Case Case 22 I1 Install Install anan independent independent $43,639

$43,639 $261,832

$261,832 $5,800,000

$5,800,000 $50,320

$50,320* $301,920

$301,920 $59,355

$59,355 $356,129

$356,129 method method of of suppression suppression pool pool cooling.

cooling.

22 Install Install aafiltered filtered containment containment $0

$0 $0

$0 $3,000,000

$3,000,000 $0

$0 $0

$0 $0

$0 $0

$0 vent to provide vent to providefission fission product product scrubbing.

scrubbing. OptionOption 1:1:Gravel Gravel Bed Bed*Filter Filter Option Option 2: 2: Multiple MultipleVenturi Venturi Scrubber Scrubber 33 Install Installaa containment containmentvent vent $10,283

$10,283 $61,701

$61,701 >$2,000,000

>$2,000,000 $11,702

$11,702 $70,211

$70,211 $14,207

$14,207 $85,244

$85,244 large enough large enough to toremove remove ATWS ATWSdecay decay heat.

heat.

44 Create Createaalargelargeconcrete concrete $436,759

$436,759 $2,620,551

$2,620,551 >$100

>$100million million $492,136

$492,136 $2,952,813

$2,952,813 $610,307

$610,307 $3,661,845

$3,661,845 crucible with crucible with heat heatremoval removal -

potential potentialunder underthe the basemat basemat totocontain contain molten molten core coredebris.

debris.

55 Create Createaawater-cooled water-cooled rubble rubble $436,759

$436,759 $2,620,551

$2,620,551 $19,000,000

$19,000,000 $498,057

$498,057 $2,988,339

$2,988,339 $610,307

$610,307 $3,661,845

$3,661,845 bed bed on on the thepedestal.

pedestal.

E.2-45 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 Pilgrim Nuclear Nuclear Power Power Station Applicant's Applicant's Environmental Environmental Report Operating Operating License Renewal Renewal Stage Table Table E.2-2E.2-2 Sensitivity Sensitivity Analysis Results (Continued)

(Continued)

Upper Upper Upper Upper Upper Upper Phase Phase Estimated Estimated Bound Bound Estimated Estimated Bound Bound Estimated Estimated Bound Bound IIII Benefit Benefit Estimated Estimated Estimated Estimated Benefit

.Benefit Estimated Estimated Benefit Benefit Estimated Estimated SAMA SAASAMA SAMA Benefit Benefit CotBenefit Benefit Benefit Benefit Cost ID 10 Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Base.Line BsLie Base Line BeLieCase Case I1 Case I1 Case Case Case 22 Case Case 22 66 Provide Provide modification modification for for $2,153

$2,153 $12,915

$12,915 >$1,000,000

>$1,000,000 $2,425

$~,425 $14,551

$14,551 $3,008 $18,048 flooding flooding the the drywell drywell head head 77 Enhance Enhance firefire protection protection $10,763

$10,763 $64,577

$64,577 >$2,500,000

>$2,500,000 $12,127

$12,127 $72,764 $15,040 $90,238 system system and/or and/or SGTS SGTS hardware and procedures.

hardware and procedures.

88 Create Create aa core core melt melt source source $436,759

$436,759 $2,620,551

$2,620,551 >$5,000,000

>$5,000,000 $498,057

$498,057 $2,988,339

$2,988,339 $610,307

$610,307 $3,661,845

$3,661,845 reduction system.

reduction system.

99 Install Install a a passive passive containment containment $44,037

$44,037 $264,219

$264,219 $5,800,000

$5,800,000 $50,845

$50,845 $305,069 $59,803

$59,803 $358,816

$358,816 spray system.,

spray system.

10 10 Strengthen Strengthen primary/

primaryl $205,571

$205,571 $1,233,428

$1,233,428 $12,000,000

$12,000,000 $231,636

$231,636 $1,389,815

$1,389,815 $287,257

$287,257 $1,723,540

$1,723,540 secondary secondary containment.

containment.

11 11 Increase Increase the the depth depth of ofthe the $4,305

$4,305 $25,831

$25,831 >$5,000,000

>$5,000,000 $4,851

$4,851 $29,105

$29,105 $6,016

$6,016 $36,095

$36,095 concrete basemat or concrete base mat or use use an an alternative alternative concrete concrete material material toto ensure ensure melt-through melt-through does does not occur not occur 12 12 Provide Provide aa reactor reactorvessel vessel $3,229

$3,229 $19,373

$19,373 $2,500,000

$2,500,000 $3,638

$3,638 $21,828

$21,828 $4,512

$4,512 $27,071

$27,071 exterior cooling exterior cooling system system (see (see

1. 7)

1. 7) _

E.2-46 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 Power Station Applicant's Environmental Environmental Report Report License Renewal Operating license Renewal Stage Table Table E.2-2 E.2-2 Sensitivity Analysis Results Sensitivity Analysis Results (Continued)

(Continued)

Upper Upper Upper Upper Upper Upper Phase Phase Estimated Estimated Bound Bound Estimated Estimated Bound Bound Estimated Estimated Bound Bound II Benefit Benefit Estimated Estimated Benefit Benefit Estimated Estimated Benefit Benefit Estimated Estimated II Estimated SAMA SAMA SAMA SAMA Benefit Benefit Cost Cost Benefit Benefit Benefit Benefit ID 10 L Sensitivity Sensitivity Sensitivity Sensitivity r;- Sensitivity Sensitivity Sensitivity Sensitivity Base Baseline Base Line Case I Case I Case Case 1 ,-

Case 1 Case 2 2 Case Case 2 2 13 13 Construct Construct aa building building to to be be $10,763

$10,763 $64,577

$64,577 >$2,000,000

>$2,000,000 $12,273

$12,273 $73,640

$73,640 $15,040

$15,040 $90,238

$90,238 connected to connected to primary/

primaryl secondary secondary containment containmentthat that is is maintained maintained at at aa vacuum vacuum 14 14 2.g.

2.g. Dedicated Dedicated Suppression Suppression $43,639

$43,639 $261,832

$261,832 $5,800,000

$5,800,000 $51,067

$51,067 $306,400

$306,400 $59,355

$59,355 $356,129

$356,129 Pool Pool Cooling Cooling 15 15 3.a.

3.a. Create Create a a larger largervolume volume in in $205,571

$205,571 $1,233,428

$1,233,428 $8,000,000

$8,000;000 $234,423

$234,423 $1,406,537

$1,406,537 $287,257

$287,257 $1,723,540

$1,723,540 containment.

containment.

16 16 3.b.

3.b. Increase Increase containment containment $205,571

$205,571 $1,233,428

$1,233,428 $12,000,000

$12,000,000 $234,423

$234,423 $1,406,537

$1,406,537 $287,257

$287,257 $1,723,540

$1,723,540 pressure pressure capability capability (sufficient (sufficient "

pressure pressure to towithstand withstand severe severe accidents).

accidents ).

17 17 3.c.

3.c. Install Install improved improved vacuum vacuum $0

$0 $o

$0 >$1,000,000

>$1,000,000 $0

$0

$0

$0 $0

$0 $0

$0 breakers breakers (redundant (redundantvalves valves inin each line).

each line).

18 18 3.d.

3.d. Increase Increase the thetemperature temperature $2,153

$2,153 $12,915

$12,915 $12,000,000

$12,000,000 $2,455

$2,455 $14,728

$14,728 $3,008

$3,008 $18,048

$18,048 margin for seals.

margin for seals.

19 19 5.b/c.

5.b/c. install Installaa filtered filtered vent vent $0

$0 $0

$0 $3,000,000

$3,000,000 $0

$0 $0

$0 $0

$0 $0

$0 E.2-47 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 Environmental Report Operating License Renewal Stage Table E.2-2 E.2-2 Sensitivity Analysis Results (Continued)

(Continued)

Upper Upper Upper Upper Upper Upper Phase Phase Estimated Estimated Bound Bound Estimated Estimated Bound Bound Estimated Estimated Bound Bound 11 II Benefit Benefit Estimated Estimated Estimated Estimated Benefit Benefit Estimated Estimated Benefit Benefit Estimated Estimated SAMA Benefit Benefit Benefit SAMA SAMA Benefit Cost Cost Benefit Benefit 10 ID ..

Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Base Line Base Line Case 1I Case I1 Case 2 Case 2 20 7.a. Provide aa method 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 13.a. altemate method of $10,763

$10,763 $64,577

$64,577 >$2,500,000 $12,273 $73,640 $15,040 $90,238 reactor building spray.

22 14.a. Provide a means of 14.a. $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

• • 25 Provide additional DC Provide DC battery battery $24,393 $146,356 $500,000

$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.

acid 27 Modification for Improving DC Modification $19,761 $118,568 $500,000 $23,377 $140,262 $26,044

.$26,044 $156,263 Bus Reliability 28 2.i. Provide 16-hour 2.1. 16-hour SBO $24,393 $146,356 $500,000 $28,207 $169,242 $33,598 $201,588 injection.

E.2-48 E.2-48

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

(Continued)

Upper Upper Upper Upper Upper Upper Phase Phase Estimated Estimated Bound Bound Estimated Estimated Bound Bound Estimated Estimated Bound Bound IP Benefit Benefit Estimated Estimated Estimated Benefit Benefit Estimated Estimated Benefit Benefit Estimated Estimated II Estimated SSAMA, SAMA. Benefit Benefit Cost Benefit Benefit SAMA Cost Benefit Benefit ID 10 Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Base Base Line Line Base Base Line Line Case I - Case 1 Case 2 Case 2

, Case 1 Case 1 Case 2 Case 2 29 29 9.b.

9.b. Provide Provide an an alternate alternate $44,281

$44,281 $265,687

$265,687 >$2,000,000

>$2,000,000 $50,546

$50,546 $303,278

$303,278 $60,956

$60,956 $365,738

$365,738 pump power pump power source.

source.

30 30 9.g.

9.g. AC AC Bus Bus Cross-Ties Cross*lies $78,902

$78,902 $473,410

$473,410 $146,120

$146,120 $91,662

$91,662 $549,972

$549,972 $106,357

$106,357 $638,142

$638,142 31 31 10.a.

10.a. Add Add aa dedicated dedicated DC DC $150,504

$150,504 $903,025

$903,025 $3,000,000

$3,000,000 $178,405

$178,405 $1,070,432

$1,070,432 $201,864

$201,864 $1,211,183

$1,211,183 power supply.

power supply.

32 32 10.b.

10.b. Install Install additional additional $150,504

$150,504 $903,025

$903,025 $3,000,000

$3,000,000 $178,405

$178,405 $1,070,432

$1,070,432 $201,864

$201,864 $1,211,183

$1,211,183 batteries batteries or ordivisions.

divisions.

33 33 10.c.

10.c. Install Install fuel fuel cells.

cells. $24,393

$24,393 $146,356

$146,356 >$2,000,000

>$2,000,000 $28,207

$28,207 $169,242

$169,242 $33,598

$33,598 $201,588

$201,588 34 34 10.d.

10.d. DC DC Cross-Ties Cross-lies $19,761

$19,761 $118,568

$118,568 $13,000

$13,000 $23,377

$23,377 $140,262

$140,262 $26,044

$26,044 $156,263

$156,263 35 35 10.e.

10.e. Extended Extended SBO seo $24,393

$24,393 $146,356

$146,356 $500,000

$500,000 $28,207

$28,207 $169,242

$169,242 $33,598

$33,598 i' $201,588

$201,588 provisions.

provisions.

36 36 Locate Locate RHR RHR inside inside $2,749

$2,749 $16,497

$16,497 >$500,000

>$500,000 $3,213

$3,213 $19,276

$19,276 $3,680

$3,680 $22,077

$22,077 containment.

containment.

37 37 Increase Increase frequency frequencyof ofvalve valve $4,025

$4,025 $24,148

$24,148 $100,000

$100,000 $4,688

$4,688 $28,127

$28,127 $5,407

$5,407 $32,444

$32,444 leak testing.

leak testing:

38 38 8.e.

8.e. improve ImproveMSIV MSIVdesign.

design. $0

$0 $0

$0 >$2,000,000

>$2,000,000 $0

$0 $0

$0 $0

$0 $0

$0 E.2-49 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 Environmental Report Operating License Renewal Stage Operating Table E.2-2 Table E.2-2 Sensitivity Analysis Results (Continued)

Sensitivity (Continued)

Upper Upper Upper Upper Upper Upper Phase Phase Estimated Estimated Bound Bound Estimated Estimated Bound Bound Estimated Estimated .Bound Bound IIIi Benefit Benefit Estimated Estimated Estimated Benefit Estimated Estimated Benefit Estimated Estimated SAMA SAMA Benefit Benefit Benefit SAMA SAMA Benefit Cost Cost Benefit Benefit ID 10  ; L Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Base Line Base Base Line Line Case I1 Case Case Case 1I Case Case 2 2 Case Case 22 39 Install an independent independent diesel $0

$0 $0

$0 $135,000 $0

$0 $0

$0 $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

• 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 aa passive high $18,369 $110,212 >$2,000,000 $21,902 $131,415 $24,477 $146,860 pressure system.

43 2.d. Improved 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 45 8.c.

8.c. Add Add a a diverse injection diverse injection $18,369

$18,369 $110,212

$110,212 >$2,000,000

>$2,000,000 $21,902

$21,902 $131,415

$131,415 $24,477

$24,477 $146,860

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

(Continued)

Upper Upper Upper Upper Upper Upper Phase Phase Estimated Estimated Bound Bound Estimated Estimated Bound Bound Estimated Estimated Bound Bound 11 II Benefit Benefit . Estimated Estimated Estimated Estimated Benefit Benefit Estimated Estimated Benefit Benefit Estimated Estimated SSAMA SAMA Benefit Benefit Cost Benefit Benefit SAMA Cost Benefit Benefit ID 10 Sensitivity Sensitivity Sensitivity Sensitivity SensitivIty Sensitivity Sensitivity Sensitivity Base Line BaseLine Base Base Line Line Case Case I1 Case Case I1 Case Case 2 2 Case Case 2 2 48 48 Diversify Diversify explosive explosive valve valve $2,153

$2,153 $12,915

$12,915 >$200,000

>$200,000 $2,425

$2,425 $14,551

$14,551 $3,008

$3,008 < $18,048

$18,048 operation.

operation.

49 49 Increase Increase the the reliability reliability of of $5,300

$5,300 $31,799

$31,799 >$1,500,000

>$1,500,000 $6,163

$6,163. $36,978

$36,978 $7,135

$7,135 $42,811

$42,811 SRVs SRVs by by adding adding signals signals toto open open them them automatically.

automatically.

50 50 8.e.

8.e. Improve Improve SRV SRV design.

design. $32,396

$32,396 $194,378

$194,378 >$2,000,000

>$2,000,000 $37,767

$37,767 $226,602

$226,602 $43,483

$43,483 $260,897

$260,897 51 51 Provide Provide self-cooled self-cooled ECCS ECCS $4,902

$4,902 $29,412

$29,412 >$200,000

>$200,000 $5,638

$5,638 $33,829

$33,829 $6,687

$6,687 $40,125

$40,125 pump pump seals.

seals.

52 52 Provide Provide digital digital large large break break $2,352

$2,352 $14,109

$14,109 >$1 00,000

>$100,000 $2,688

$2,688 $16,126

$16,126 $3,232 $19,391

$19,391 LOCA LOCA protection.

protection .

53 53 Control Control containment containment venting venting $22,873

$22,873 $137,237

$137,237 $300,000

$300,000 $26,653

$26,653 $159,919

$159,919 $30,716

$30,716 $184,299

$184,299 within within a a narrow narrow band band ofof pressure pressure 54 54 Install Install a a bypass bypass switch switch to to $3,627

$3,627 $21,761

$21,761 $1,000,000

$1,000,000 $4,163

$4,163 $24,978

$24,978 $4,960

$4,960 $29,758

$29,758 bypass bypass thethe low low reactor reactor pressure pressure interlocks interlocks ofof LPCI LPCI oror core core spray spray injection injection valves.

valves.

55 55 Improve Improve SSW SSW System System andand $59,385

$59,385 $356,310

$356,310 >$5

>$5 million million $67,986

$67,986 $407,918

$407,918 $81,467

$81,467 $488,799 RBCCW RBCCW pump pump recovery.

recovery.

E.2-51 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 Applicant's Environmental Report Operating Operating License Renewal Renewal Stage Table E.2-2 E.2-2 Sensitivity Analysis Results (Continued)

(Continued)

Upper Upper Upper Upper Upper Upper Phase Phase Estimated Estimated Bound Bound Estimated Estimated Bound Bound Estimated Estimated Bound Bound IIII Benefit Estimated Estimated Estimated Estimated Benefit Benefit Estimated Estimated Benefit Benefit Estimated Estimated SAMA Benefit Benefit Benefit SAMA SAMA SAMA Benefit Cost Cost Benefit Benefit ID 10 Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Sensitivity Base Line Base Line Baseline Base Line Case Case I1 Case 1I Case Case 22 Case Case 22 Case 56 Provide redundant DC power $36,773 $220,639 $112,400

$112,400 $43,541

$43,541 $261,247 $48,408 $290,449 supplies to DTV valves.

57 Proceduralize Proceduralize the the use use of of $29,213 $175,279

$175,279 $26,000 $33,568

$33,568 $201,406

$201,406 $39,901 $239,406 diesel fire diesel fire pump pump hydroturbine hydroturbine in in the event of EDG A failure or unavailability.

58 Proceduralize the Proceduralize the operator operator $31,799

$31,799 $190,797 $50,000

$50,000 $36,980

$36,980 $221,878 $42,811 $256,868 action to feed action feed 81 B1loads via 83

'pads via B3 When A5 AS isis unavailable post-trip.

59 Provide redundant path from $154,966 $929,797 $1,956,000

$1,956,000 $176,682 $1,060,091 $213,620

$213,620 $1,281,720

$1,281,720 fire protection fire protection pump pump discharge to lPClloops LPCI loops AA and B 8 cross-tie.

E.2-52 E.2-52