PNP 2014-074, Relief Request Number RR 4-20 Proposed Alternative, Use of Alternate ASME Code Case N-716

Relief Request Number RR 4-20 Proposed Alternative, Use of Alternate ASME Code Case N-716

ML14226A618
Person / Time
Site:	Palisades  (DPR-020)
Issue date:	08/14/2014
From:	Hardy J Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PNP 2014-074
Download: ML14226A618 (89)

Text

Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant t

27780 Blue Star Memorial Highway Covert, Ml 49043-9530 Tel 269 764 2000 Jeffery A. Hardy Regulatory Assurance Manager PNP 2014-074 August 14, 2014 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Relief Request Number RR 4-20 Proposed Alternative, Use of Alternate ASME Code Case N-716 Palisades Nuclear Plant Docket 50-255 License No. DPR-20

Dear Sir or Madam:

Pursuant to 10 CFR 50.55a(a)(3)(i), Entergy Nuclear Operations, Inc. (ENO) hereby requests NRC approval of the Request for Relief number RR 4-20 for a proposed alternative for the Palisades Nuclear Plant (PNP). This alternative is for the current fourth 10-year lSI interval.

This request is associated with the use of an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Code Case N-716, Alternative Piping Classification and Examination RequirementsSection XI, Division 1, a risk informed, safety-based inservice (ISI) program. The information provided in the enclosed request demonstrates that the proposed alternative provides an acceptable level of quality and safety.

ENO requests NRC approval by August 14, 2015 to support plans to implement the proposed alternative during the fourth ten-year interval, third period.

Summary of Commitments This letter contains no new commitments and no revised commitments.

Sincerely, jah/jpm

~Entergy PNP 2014-074 August 14, 2014 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043-9530 Tel 269 764 2000 Jeffery A. Hardy Regulatory Assurance Manager

SUBJECT:

Relief Request Number RR 4 Proposed Alternative, Use of Alternate ASME Code Case N-716 Palisades Nuclear Plant Docket 50-255 License No. DPR-20

Dear Sir or Madam:

Pursuant to 10 CFR 50.55a(a)(3)(i), Entergy Nuclear Operations, Inc. (END) hereby requests NRC approval of the Request for Relief number RR 4-20 for a proposed alternative for the Palisades Nuclear Plant (PNP). This alternative is for the current fourth 10-year lSI interval.

This request is associated with the use of an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Code Case N-716, Alternative Piping Classification and Examination RequirementsSection XI, Division 1, a risk informed, safety-based inservice (lSI) program. The information provided in the enclosed request demonstrates that the proposed alternative provides an acceptable level of quality and safety.

END requests NRC approval by August 14, 2015 to support plans to implement the proposed alternative during the fourth ten-year interval, third period.

Summary of Commitments This letter contains no new commitments and no revised commitments.

Sincerely, jah/jpm

PNP 201 4-074 Page 2 of 2

Enclosure:

Proposed Alternative cc:

Administrator, Region Ill, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC PNP 2014-074 Page 2 of 2

Enclosure:

Proposed Alternative cc:

Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC

ENCLOSURE ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

PALISADES NUCLEAR PLANT 10 CFR 50.55a Relief Request Number RR 4-20 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(i)

Alternative Provides Acceptable Level of Quality and Safety Risk-Informed/Safety Based Inservice Inspection Alternative for Class 1 and 2 Piping 30 pages follow ENCLOSURE ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

PALISADES NUCLEAR PLANT 10 CFR 50.55a Relief Request Number RR 4-20 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(i)

Alternative Provides Acceptable Level of Quality and Safety Risk-Informed/Safety Based Inservice Inspection Alternative for Class 1 and 2 Piping 30 pages follow

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Plant Site -

Palisades Nuclear Power Plant (PNP)

Unit:

Interval -

Fourth Ten Year Interval

Dates:

Third Period: May 13, 2012 to December 12, 2015 Requested Approval is requested by August 15, 2015 Date for Approval:

ASME Code All Class 1 and 2 piping welds Examination Categories B-F, B-J, C-F-i, Components and C-F-2.

Affected:

Applicable For PNP, the 4 th ten year Inservice Inspection (ISI) Interval began on Code Edition December 13, 2006. The applicable Code of Record is the American and Addenda:

Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2001 Edition through the 2003 Addenda.

Applicable The requirements from which an alternative is requested are specified in Code the ASME Code, Section Xl, 2001 Edition with the 2003 Addenda, Requirements:

IWB-2200, IWB-2420, IWB-2430, IWB-2500, Table IWB-2500-i, (Examination Categories B-F and B-J); and in IWC-2200, IWC-2420, IWC-2430, and IWC-2500, Table IWC-2500-1 (Examination Categories C-F-i and C-F-2).

Reason for The objective of this submittal is to request the use of a risk-Request:

informed/safety based (RISB) ISI process for the inservice inspection of Class 1 and 2 piping.

Proposed In lieu of the ASME Code requirements, Entergy Nuclear Operations, Inc.

Alternative (ENO), proposes to use a RIS_B process at PNP as an alternate to the and Basis for ASME Section Xl ISI program for Class 1 and 2 piping. The RIS_B Use:

process used in this submittal is based upon ASME Code Case N-7i6, Alternative Piping Classification and Examination Requirements, Section Xl, Division 1.

Code Case N-716 is founded, in large part, on the RI-ISI process described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657, Rev. B-A, Revised Risk-Informed Inservice Inspection Evaluation Procedure, December 1999 (ADAMS Accession No. ML013470102) which was previously reviewed and approved by the U.S.

Nuclear Regulatory Commission (NRC).

Page 1 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Plant Site-Palisades Nuclear Power Plant (PNP)

Unit:

Interval-Fourth Ten Year Interval-Dates:

Third Period: May 13, 2012 to December 12, 2015 Requested Approval is requested by August 15, 2015 Date for Approval:

ASME Code All Class 1 and 2 piping welds - Examination Categories B-F, B-J, C-F-1, Components and C-F-2.

Affected:

Applicable For PNP, the 4th ten year Inservice Inspection (lSI) Interval began on Code Edition December 13, 2006. The applicable Code of Record is the American and Addenda:

Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components,2001 Edition through the 2003 Addenda.

Applicable The requirements from which an alternative is requested are specified in Code the ASME Code,Section XI, 2001 Edition with the 2003 Addenda, Requirements:

IWB-2200, IWB-2420, IWB-2430, IWB-2500, Table IWB-2500-1, (Examination Categories B-F and B-J); and in IWC-2200, IWC-2420, IWC-2430, and IWC-2500, Table IWC-2500-1 (Examination Categories C-F-1 and C-F-2).

Reason for The objective of this submittal is to request the use of a risk-Request:

informed/safety based (RIS_B) lSI process for the inservice inspection of Class 1 and 2 piping.

Proposed In lieu of the ASME Code requirements, Entergy Nuclear Operations, Inc.

Alternative (END), proposes to use a RIS_B process at PNP as an alternate to the and Basis for ASME Section XI lSI program for Class 1 and 2 piping. The RIS_B Use:

process used in this submittal is based upon ASME Code Case N-716, Alternative Piping Classification and Examination Requirements,Section XI, Division 1.

Code Case N-716 is founded, in large part, on the RI-ISI process described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657, Rev. B-A, Revised Risk-Informed Inservice Inspection Evaluation Procedure, December 1999 (ADAMS Accession No. ML013470102) which was previously reviewed and approved by the U.S.

Nuclear Regulatory Commission (NRC).

Page 1 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

In general, a risk-informed program replaces the number and locations of nondestructive examination (NDE) inspections based on ASME Code,Section XI requirements with the number and locations of these inspections based on the risk-informed guidelines. The risk-informed program is consistent with the concept that, by focusing inspections on the most safety-significant welds, the number of inspections can be reduced while at the same time maintaining protection of public health and safety.

The NRC-approved EPRI TR 112657, Rev. B-A, includes steps which, when successfully applied, satisfy the guidance provided in Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis and RG 1.178, An Approach For Plant-Specific Risk-Informed Decision Making for Inservice Inspection of Piping. These steps are:

Scope definition Consequence evaluation Degradation mechanism evaluation Piping segment definition Risk categorization lnspection/NDE selection Risk impact assessment Implementation monitoring and feedback These same steps were also applied to this RIS_B process and it is concluded that this RIS_B process alternative also meets the intent and principles of RG 1.174 and RG 1.178.

In general, the methodology in Code Case N-716 replaces a detailed evaluation of the safety significance of each pipe segment required by EPRI TR 112657, Rev. B-A with a generic population of high safety-significant segments, supplemented with a rigorous flooding analysis to identify any plant-specific high safety-significant segments (Class 1, 2, 3, or Non-Class). The flooding analysis was performed in accordance with RG 1.200 and ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications.

By using risk insights to focus examinations on more safety-significant locations, while meeting the intent and principles of RG 1.174 and RG 1.178, this proposed RIS_B program will continue to maintain an acceptable level of quality and safety. Additionally, all piping components, regardless of risk classification, will continue to receive ASME Code-required pressure testing, as part of the current ASME Code, Section Xl program. Therefore, approval for this alternative to the requirements of IWB-2200, IWB-2420, IWB-2430, and IWB-2500, Table IWB-2500-1, (Examination Categories B-F and B-J), and IWC-2200, IWC-2420, IWC-2430, IWC-2500, Table IWC-2500- 1 (Examination Page 2 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

In general, a risk-informed program replaces the number and locations of nondestructive examination (NDE) inspections based on ASME Code,Section XI requirements with the number and locations of these inspections based on the risk-informed guidelines. The risk-informed program is consistent with the concept that, by focusing inspections on the most safety-significant welds, the number of inspections can be reduced while at the same time maintaining protection of public health and safety.

The NRC-approved EPRI TR 112657, Rev. B-A, includes steps which, when successfully applied, satisfy the guidance provided in Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis and RG 1.17B, An Approach For Plant-Specific Risk-Informed Decision Making for Inservice Inspection of Piping. These steps are:

Scope definition Consequence evaluation Degradation mechanism evaluation Piping segment definition Risk categorization InspectionlNDE selection Risk impact assessment Implementation monitoring and feedback These same steps were also applied to this RIS_B process and it is concluded that this RIS_B process alternative also meets the intent and principles of RG 1.174 and RG 1.17B.

In general, the methodology in Code Case N-716 replaces a detailed evaluation of the safety significance of each pipe segment required by EPRI TR 112657, Rev. B-A with a generic population of high safety-significant segments, supplemented with a rigorous flooding analysis to identify any plant-specific high safety-significant segments (Class 1,2,3, or Non-Class). The flooding analysis was performed in accordance with RG 1.200 and ASMEI ANS RA-Sa-2009, Addenda to ASMEIANS RA-S-200B, Standard for Level11Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications.

By using risk insights to focus examinations on more safety-significant locations, while meeting the intent and principles of RG 1.174 and RG 1.17B, this proposed RIS_B program will continue to maintain an acceptable level of quality and safety. Additionally, all piping components, regardless of risk classification, will continue to receive ASME Code-required pressure testing, as part of the current ASME Code,Section XI program. Therefore, approval for this alternative to the requirements of IWB-2200, IWB-2420, IWB-2430, and IWB-2500, Table IWB-2500-1, (Examination Categories B-F and B-J), and IWC-2200, IWC-2420, IWC-2430, IWC-2500, Table IWC-2500-1 (Examination Page 2 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Categories C-F-i and C-F-2), is requested in accordance with 10 CFR 50.55a(a)(3)(i). A Palisades specific template for the application of ASME Code Case N-716 is attached.

All other ASME Code,Section XI requirements for which relief was not specifically requested in this relief request remain applicable.

Duration of Use of the proposed alternative is requested for the duration of the third Proposed period, fourth interval (May 13, 2012 to December 12, 2015).

Alternative:

Precedents:

Similar alternatives have been approved for Vogtle Electric Generating Plant, Donald C. Cook 1 and 2, Grand Gulf Nuclear Station, Waterford-3 and North Anna 1 & 2.

References:

Vogtle Electric Generating Plant Safety Evaluation

- See ADAMS Accession No. ML100610470.

D. C. Cook Safety Evaluation

- See ADAMS Accession No. ML072620553.

Grand Gulf Nuclear Station Safety Evaluation-See ADAMS Accession No. ML072430005.

Waterford-3 Safety Evaluation See ADAMS Accession No. ML080980120.

Page 3 of 30 Duration of Proposed Alternative:

Precedents:

References:

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Categories C-F-1 and C-F-2), is requested in accordance with 10 CFR SO.SSa(a)(3)(i). A Palisades specific template for the application of ASME Code Case N-716 is attached.

All other ASME Code,Section XI requirements for which relief was not specifically requested in this relief request remain applicable.

Use of the proposed alternative is requested for the duration of the third period, fourth interval (May 13, 2012 to December 12, 201S).

Similar alternatives have been approved for Vogtle Electric Generating Plant, Donald C. Cook 1 and 2, Grand Gulf Nuclear Station, Waterford-3 and North Anna 1 & 2.

Vogtle Electric Generating Plant Safety Evaluation - See ADAMS Accession No. ML100610470.

D. C. Cook Safety Evaluation - See ADAMS Accession No. ML072620SS3.

Grand Gulf Nuclear Station Safety Evaluation-See ADAMS Accession No. ML07243000S.

Waterford-3 Safety Evaluation - See ADAMS Accession No. ML080980120.

Page 3 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

TEMPLATE SUBMITTAL APPLICATION OF ASME CODE CASE N-716 RISK-INFORMED/SAFETY-BASED (RIS_B)

INSERVICE INSPECTION PROGRAM PLAN Page 4 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

TEMPLATE SUBMITTAL APPLICATION OF ASME CODE CASE N-716 RISK-INFORMED/SAFETY-BASED (RIS_B)

INSERVICE INSPECTION PROGRAM PLAN Page 4 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Technical Acronyms/Definitions Used in the Template AS Accident Sequence Analysis ASEP Accident Sequence Evaluation Program ASME American Society of Mechanical Engineers BER Break Exclusion Region BLDS Blowdown System CAFTA Computer-Aided Fault Tree Analysis CC PRA abbreviation for Capacity Category CC Crevice Corrosion CCDP Conditional Core Damage Probability CCF Common Cause Failure CCW Component Cooling Water CDF Core Damage Frequency CIV Containment Isolation Valve CLERP Conditional Large Early Release Probability CSS Containment Spray System CVCS Chemical Volume and Control System DA Data analysis DM Degradation Mechanism E-C Erosion-Corrosion ECSCC External Chloride Stress Corrosion Cracking EOOS Equipment Out of Service ESS Engineered Safeguards System FAC Flow-Accelerated Corrosion F&O Facts and Observations FLB Feedwater Line Break FPS Fire Protection System FT Fault tree FWS Feedwater System HELB High Energy Line Break HEP Human Error Probability HFE Human Failure Event HR Human Reliability HRA Human Reliability Analysis HSS High Safety-Significant lE Initiating Events Analysis IF Internal Flooding IFIV Inside First Isolation Valve IGSSC Intergranular Stress Corrosion Cracking ILOCA Isolable Loss of Coolant Accident IPE Individual Plant Evaluation LE LERF Analysis LERF Large Early Release Frequency LOCA Loss of Coolant Accident LOOP Loss of Off-Site Power LSS Low Safety-Significant MAAP Modular Accident Analysis Program MIC Microbiologically-Influenced Corrosion Page 5 of 30 AS ASEP ASME BER BLDS CAFTA CC CC CCDP CCF CCW CDF CIV CLERP CSS CVCS DA DM E-C ECSCC EOOS ESS FAC F&O FLB FPS FT FWS HELB HEP HFE HR HRA HSS IE IF IFIV IGSSC I LOCA IPE LE LERF LOCA LOOP LSS MAAP MIC ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Technical Acronyms/Definitions Used in the Template Accident Sequence Analysis Accident Sequence Evaluation Program American Society of Mechanical Engineers Break Exclusion Region Blowdown System Computer-Aided Fault Tree Analysis PRA abbreviation for Capacity Category Crevice Corrosion Conditional Core Damage Probability Common Cause Failure Component Cooling Water Core Damage Frequency Containment Isolation Valve Conditional Large Early Release Probability Containment Spray System Chemical Volume and Control System Data analysis Degradation Mechanism Erosion-Corrosion External Chloride Stress Corrosion Cracking Equipment Out of Service Engineered Safeguards System Flow-Accelerated Corrosion Facts and Observations Feedwater Line Break Fire Protection System Fault tree Feedwater System High Energy Line Break Human Error Probability Human Failure Event Human Reliability Human Reliability Analysis High Safety-Significant Initiating Events Analysis Internal Flooding Inside First Isolation Valve Intergranular Stress Corrosion Cracking Isolable Loss of Coolant Accident Individual Plant Evaluation LERF Anaiysis Large Early Release Frequency Loss of Coolant Accident Loss of Off-Site Power Low Safety-Significant Modular Accident Analysis Program Microbiologically-Influenced Corrosion Page 5 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 5055a(a)(3)(i)

Technical Acronyms/Definitions Used in the Template (Continued)

MOV Motor Operated Valve MSS Main Steam System MU Model Update NDE Nondestructive Examination NNS Non-Nuclear Safety NPS Nominal Pipe Size PBF Pressure Boundary Failure PCP Primary Coolant Pump PCS Primary Coolant System PIT Pitting PLOCA Potential Loss of Coolant Accident POD Probability of Detection PRA Probabilistic Risk Assessment PSA Probabilistic Safety Assessment PWR Pressurized Water Reactor PWSCC Primary Water SCC QU Quantification RCP Reactor Coolant Pump RCPB Reactor Coolant Pressure Boundary RG Regulatory Guide RHRS Residual Heat Removal System RI-BER Risk-Informed Break Exclusion Region RI-ISI Risk-Informed Inservice Inspection RIS_B Risk-Informed/Safety Based Inservice Inspection RM Risk Management RPV Reactor Pressure Vessel RWS Radwaste System SBO Station Blackout SC Success Criteria SDC Shutdown Cooling SLB Steam Line Break SGTR Steam Generator Tube Rupture SSC Systems, Structures, and Components SR Supporting Requirements SWS Service Water System SXI Section Xl SY Systems Analysis TASCS Thermal Stratification, Cycling, and Striping TGSCC Transgranular Stress Corrosion Cracking TR Technical Report TT Thermal Transient VAS Vent & Air Conditioning System VOL Volumetric Page 6 of 30 MOV MSS MU NDE NNS NPS PBF PCP PCS PIT PLOCA POD PRA PSA PWR PWSCC QU RCP RCPB RG RHRS RI-BER RI-ISI RIS_B RM RPV RWS SBO SC SDC SLB SGTR SSC SR SWS SXI SY TASCS TGSCC TR TT VAS VOL ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Technical Acronyms/Definitions Used in the Template (Continued)

Motor Operated Valve Main Steam System Model Update Nondestructive Examination Non-Nuclear Safety Nominal Pipe Size Pressure Boundary Failure Primary Coolant Pump Primary Coolant System Pitting Potential Loss of Coolant Accident Probability of Detection Probabilistic Risk Assessment Probabilistic Safety Assessment Pressurized Water Reactor Primary Water SCC Quantification Reactor Coolant Pump Reactor Coolant Pressure Boundary Regulatory Guide Residual Heat Removal System Risk-Informed Break Exclusion Region Risk-Informed Inservice Inspection Risk-Informed/Safety Based Inservice Inspection Risk Management Reactor Pressure Vessel Radwaste System Station Blackout Success Criteria Shutdown Cooling Steam Line Break Steam Generator Tube Rupture Systems, Structures, and Components Supporting Requirements Service Water System Section XI Systems Analysis Thermal Stratification, Cycling, and Striping Transgranular Stress Corrosion Cracking Technical Report Thermal Transient Vent & Air Conditioning System Volumetric Page 6 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table of Contents 1.

Introduction 1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 1.2 Probabilistic Safety Assessment (PSA) Quality 2.

Proposed Alternative to Current 151 Programs 2.1 ASME Section Xl 2.2 Augmented Programs 3.

Risk-Informed/Safety-Based ISI Process 3.1 Safety Significance Determination 3.2 Failure Potential Assessment 3.3 Element and NDE Selection 3.3.1 Current Examinations 3.3.2 Successive Examinations 3.3.3 Scope Expansion 3.3.4 Program Relief Requests 3.4 Risk Impact Assessment 3.4.1 Quantitative Analysis 3.4.2 Defense-in-Depth 3.5 Implementation 3.6 Feedback (Monitoring) 4.

Proposed lSl Plan Change 5.

References/Documentation Attachment A: Consideration of the Adequacy of Probabilistic Risk Assessment Model for Application of Code Case N71 6 Palisades PRA Response to RG 1.200 Peer Review Findings Page 7 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table of Contents

1. Introduction 1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 1.2 Probabilistic Safety Assessment (PSA) Quality

2. Proposed Alternative to Current lSI Programs 2.1 ASME Section XI 2.2 Augmented Programs

3. Risk-Informed/Safety-Based lSI Process 3.1 Safety Significance Determination 3.2 Failure Potential Assessment 3.3 Element and NDE Selection 3.3.1 Current Examinations 3.3.2 Successive Examinations 3.3.3 Scope Expansion 3.3.4 Program Relief Requests 3.4 Risk Impact Assessment 3.4.1 Quantitative Analysis 3.4.2 Defense-in-Depth 3.5 Implementation 3.6 Feedback (Monitoring)

4. Proposed lSI Plan Change

5. References/Documentation Attachment A: Consideration of the Adequacy of Probabilistic Risk Assessment Model for Application of Code Case N716 Palisades PRA Response to RG 1.200 Peer Review Findings Page 7 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 1.

INTRODUCTION Palisades Nuclear Power Plant (PNP) is currently in the fourth Inservice Inspection (ISI) interval, as defined by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Section XI Code for Inspection Program B. PNP plans to implement a risk-informed/safety-based inservice inspection (RIS_B) program in the third period of the fourth interval. The third period of the fourth ISI interval began on May 13, 2012.

The ASME Section XI Code of record for the fourth ISI interval is the ASME Section Xl 2001 Edition with the 2003 Addenda for Examination Category B-F, B-J, C-F-i, and C-F-2 Class 1, 2, 3, or Non-Class piping.

The RIS_B process used in this submittal is based upon ASME Code Case N-71 6, Alternative Piping Classification and Examination Requirements, Section Xl Division 1, which is founded in large part on the Rl-lSl process as described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657, Rev. B-A, Revised Risk-Informed Inservice Inspection Evaluation Procedure.

1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 As a risk-informed application, this submittal meets the intent and principles of Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, and RG 1.178, An Approach for Plant-Specific Risk-Informed Decision making Inservice Inspection of Piping. Additional information is provided in Section 3.4.2 relative to defense-in-depth.

1.2 Probabilistic Safety Assessment (PSA) Quality The methodology in Code Case N-71 6 provides for examination of a generic population of HSS segments, supplemented with a rigorous flooding analysis to identify any plant-specific HSS segments that need to be added. Satisfying the requirement for the plant-specific analysis requires confidence that the flooding PRA is capable of successfully identifying any significant flooding contributors that are not identified in the generic population.

RG 1.200 revision 2, An Approach for Determining the Technical Adequacy of Probabillstic Risk Assessment Results for Risk-Informed Activities, was used to demonstrate that the PRA analysis is adequate to support a risk-informed application. RG 1.200 further indicates that an acceptable approach for ensuring technical adequacy is to perform a peer review.

The technical adequacy of the PNP PRA model was determined by demonstrating through peer reviews that the PNP PRA model meets the technical elements and associated supporting requirements (SRs) of the ASME PRA Standard as clarified in NRC RG 1.200, Revision 2 (for Internal Events only). The resolution to Peer Team results shown in Attachment A demonstrates that the technical adequacy of the PNP PRA model is robust.

Page8of3o ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SS8 RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

1. INTRODUCTION Palisades Nuclear Power Plant (PNP) is currently in the fourth Inservice Inspection (lSI) interval, as defined by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Section XI Code for Inspection Program B. PNP plans to implement a risk-informed/safety-based inservice inspection (RIS_B) program in the third period of the fourth interval. The third period of the fourth lSI interval began on May 13, 2012.

The ASME Section XI Code of record for the fourth lSI interval is the ASME Section XI 2001 Edition with the 2003 Addenda for Examination Category B-F, B-J, C-F-1, and C-F-2 Class 1,2,3, or Non-Class piping.

The RIS_B process used in this submittal is based upon ASME Code Case N-716, Alternative Piping Classification and Examination Requirements,Section XI Division 1, which is founded in large part on the RI-ISI process as described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657, Rev. B-A, Revised Risk-Informed Inservice Inspection Evaluation Procedure.

1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 As a risk-informed application, this submittal meets the intent and principles of Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, and RG 1.178, An Approach for Plant-Specific Risk-Informed Decision making Inservice Inspection of Piping. Additional information is provided in Section 3.4.2 relative to defense-in-depth.

1.2 Probabilistic Safety Assessment (PSA) Quality The methodology in Code Case N-716 provides for examination of a generic population of HSS segments, supplemented with a rigorous flooding analysis to identify any plant-specific HSS segments that need to be added. Satisfying the requirement for the plant-specific analysis requires confidence that the flooding PRA is capable of successfully identifying any significant flooding contributors that are not identified in the generic population.

RG 1.200 revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, was used to demonstrate that the PRA analysis is adequate to support a risk-informed application. RG 1.200 further indicates that an acceptable approach for ensuring technical adequacy is to perform a peer review.

The technical adequacy of the PNP PRA model was determined by demonstrating through peer reviews that the PNP PRA model meets the technical elements and associated supporting requirements (SRs) of the ASME PRA Standard as clarified in NRC RG 1.200, Revision 2 (for Internal Events only). The resolution to Peer Team results shown in Attachment A demonstrates that the technical adequacy of the PNP PRA model is robust.

Page 8 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Numerous RI-ISI evaluations concluded external events are not likely to impact the consequence ranking. This position is supported by Section 2 of EPRI Report 1021467, Nondestructive Evaluation: Probabillstic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs, which concludes that quantification of these events will not change the conclusions derived from the RI-ISI process. As a result, there is no need to further consider these events.

2.

PROPOSED ALTERNATIVE TO CURRENT ISI PROGRAMS 2.1 ASME Section XI ASME Section XI Examination Categories B-F, B-J, C-F-i, and C-F-2 currently contain requirements for the nondestructive examination (NDE) of Class 1 and 2 piping components.

The alternative RIS_B Program for piping is described in Code Case N-7i6.

The RIS_B Program will be substituted for the current program for Class 1 and 2 piping (Examination Categories B-F, B-J, C-F-i and C-F-2) in accordance with 10 CFR 50.55a(a)(3)(i) by alternatively providing an acceptable level of quality and safety. Other non-related portions of the ASME Section Xl Code will be unaffected.

2.2 Augmented Programs The impact of the RIS_B application on the various plant augmented inspection programs listed below were considered. This section documents only those plant augmented inspection programs that address common piping with the RISB application scope (i.e., Class 1, 2 and 3 piping).

The plant augmented inspection program for high energy line breaks is not changed by the RIS_B Program.

The plant augmented inspection program for flow accelerated corrosion per Generic Letter (GL) 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning, is relied upon to manage this damage mechanism but is not otherwise affected or changed by the RIS_B Program.

The plant augmented inspection program for localized corrosion per Generic Letter 89-i 3, Service Water System Problems Affecting Safety-Related Equipment, is relied upon to manage localized corrosion damage mechanisms. Non-class piping in the Fire Protection System was determined to be high safety significant (CDF>i E-6 based on internal flooding results). While the sampling percentages of Code Case N-71 6 will be applied to this piping, it will be inspected under the existing effective localized corrosion program, per Section 3.6.7 of EPRI TR-i 12657.

Page 9 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Numerous RI-ISI evaluations concluded external events are not likely to impact the consequence ranking. This position is supported by Section 2 of EPRI Report 1021467, Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs, which concludes that quantification of these events will not change the conclusions derived from the RI-ISI process. As a result, there is no need to further consider these events.

2.

PROPOSED ALTERNATIVE TO CURRENT lSI PROGRAMS 2.1 ASME Section XI ASME Section XI Examination Categories B-F, B-J, C-F-1, and C-F-2 currently contain requirements for the nondestructive examination (NDE) of Class 1 and 2 piping components.

The alternative RIS_B Program for piping is described in Code Case N-716.

The RIS_B Program will be substituted for the current program for Class 1 and 2 piping (Examination Categories B-F, B-J, C-F-1 and C-F-2) in accordance with 10 CFR 50.55a(a)(3)(i) by alternatively providing an acceptable level of quality and safety. Other non-related portions of the ASME Section XI Code will be unaffected.

2.2 Augmented Programs The impact of the RIS_B application on the various plant augmented inspection programs listed below were considered. This section documents only those plant augmented inspection programs that address common piping with the RIS_B application scope (Le., Class 1, 2 and 3 piping).

The plant augmented inspection program for high energy line breaks is not changed by the RIS_B Program.

The plant augmented inspection program for flow accelerated corrosion per Generic Letter (GL) 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning, is relied upon to manage this damage mechanism but is not otherwise affected or changed by the RIS_B Program.

The plant augmented inspection program for localized corrosion per Generic Letter 89-13, Service Water System Problems Affecting Safety-Related Equipment, is relied upon to manage localized corrosion damage mechanisms. Non-class piping in the Fire Protection System was determined to be high safety significant (CDF> 1 E-6 based on internal flooding results). While the sampling percentages of Code Case N-716 will be applied to this piping, it will be inspected under the existing effective localized corrosion program, per Section 3.6.7 of EPRI TR-112657.

Page 9 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

A plant augmented inspection program has been implemented at PNP in response to Code Case N-770-1 (1), Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guidelines. The requirements of Code Case N-770-1 will be used for the inspection and management of PWSCC susceptible welds and will supplement the RIS_B Program selection process. The RIS_B Program will not be used to eliminate any Code Case N-770-1 requirements. Alloy 82/182 welds susceptible to no degradation mechanism or PWSCC only per the RIS_B Program failure potential assessment will be removed from the RIS_B population subject to element selection, and will be inspected and managed per the requirements of Code Case N-770-1 (1)(2)

Notes:

(1) Effective July 21, 2011, 10 CFR 50.55a was amended via rulemaking to incorporate by reference Code Case N7701, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1, which replaces MRP-139 for the inspection and management of PWSCC susceptible welds. PNP is managing Alloy 82/1 82 welds per the requirements of Code Case N-770-1.

(2) Alloy 82/182 welds subject to PWSCC and an additional degradation mechanism (or mechanisms) remain in the RIS_B population subject to element selection.

PNP has conducted an evaluation in accordance with MRP-1 46, Materials Reliability Program: Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines, and these results have been incorporated into the RISB Program.

3.

RISK-INFORMED/SAFETY-BASED ISI PROCESS The process used to develop the RIS_B Program conformed to the methodology described in Code Case N-71 6 and consisted of the following steps:

Safety Significance Determination (see Section 3.1)

Failure Potential Assessment (see Section 3.2)

Element and NDE Selection (see Section 3.3)

Risk Impact Assessment (see Section 3.4)

Implementation Program (see Section 3.5)

Feedback (Monitoring) (see Section 3.6)

Each of these six steps is discussed below:

Page 10 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a{a){3){i)

A plant augmented inspection program has been implemented at PNP in response to Code Case N-770-1 (1), Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guidelines. The requirements of Code Case N-770-1 will be used for the inspection and management of PWSCC susceptible welds and will supplement the RIS_B Program selection process. The RIS_B Program will not be used to eliminate any Code Case N-770-1 requirements. Alloy 82/182 welds susceptible to no degradation mechanism or PWSCC only per the RIS_B Program failure potential assessment will be removed from the RIS_B population subject to element selection, and will be inspected and managed per the requirements of Code Case N-770-1 (1)(2).

Notes:

(1) Effective July 21,2011, 10 CFR SO.SSa was amended via rulemaking to incorporate by reference Code Case N-770-1, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1, which replaces MRP-139 for the inspection and management of PWSCC susceptible welds. PNP is managing Alloy 821182 welds per the requirements of Code Case N-770-1.

(2) Alloy 821182 welds subject to PWSCC and an additional degradation mechanism (or mechanisms) remain in the RIS_B population subject to element selection.

PNP has conducted an evaluation in accordance with MRP-146, Materials Reliability Program: Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines, and these results have been incorporated into the RIS_B Program.

3.

RISK-INFORMED/SAFETY-BASED lSI PROCESS The process used to develop the RIS_B Program conformed to the methodology described in Code Case N-716 and consisted of the following steps:

Safety Significance Determination (see Section 3.1)

Failure Potential Assessment (see Section 3.2)

Element and NDE Selection (see Section 3.3)

Risk Impact Assessment (see Section 3.4)

Implementation Program (see Section 3.S)

Feedback (Monitoring) (see Section 3.6)

Each of these six steps is discussed below:

Page 10 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3.1 Safety Significance Determination The systems assessed in the RIS_B Program are provided in Table 3.1. The piping and instrumentation diagrams and additional plant information, including the existing plant ISI Program were used to define the piping system boundaries. Per Code Case N-716 requirements, piping welds are assigned safety-significance categories, which are then used to determine the examination treatment requirements. High safety-significant (HSS) welds are determined in accordance with the requirements below. Low safety-significant (LSS) welds include all other Class 2, 3, or Non-Class welds.

(1)

Class 1 portions of the reactor coolant pressure boundary (RCPB), except as provided in 10 CFR 50.55a(c)(2)(i) and (c)(2)(ii)

(2)

Applicable portions of the shutdown cooling pressure boundary function. That is, Class 1 and 2 welds of systems or portions of systems needed to utilize the normal shutdown cooling flow path either:

(a) As part of the RCPB from the reactor pressure vessel (RPV) to the second isolation valve (i.e., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds; or (b) Other systems or portions of systems from the RPV to the second isolation valve (i.e., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds (3)

That portion of the Class 2 feedwater system [greater than 4 inch nominal pipe size (NPS)] of pressurized water reactors (PWRs) from the steam generator to the outer containment isolation valve, (4)

Piping within the break exclusion region (BER) greater than 4 inch NPS for high-energy piping systems as defined by the Owner. Per Code Case N-71 6, this may include Class 3 or Non-Class piping.

(5)

Any piping segment whose contribution to core damage frequency (CDF) is greater than 1 E-06 [and per NRC feedback on the Grand Gulf and D. C. Cook RIS_B applications 1 E-07 for large early release frequency (LERF)] based upon a plant-specific PSA of pressure boundary failures (e.g., pipe whip, jet impingement, spray, inventory losses). This may include Class 3 or Non-Class piping.

3.2 Failure Potential Assessment Failure potential estimates were generated utilizing industry failure history, plant-specific failure history, and other relevant information. These failure estimates were determined using the guidance provided in NRC approved EPRI TR-1 12657 (i.e., the EPRI RI-ISI methodology), with the exception of the deviation discussed below.

Page 11 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a{a){3){i) 3.1 Safety Significance Determination The systems assessed in the RIS_B Program are provided in Table 3.1. The piping and instrumentation diagrams and additional plant information, including the existing plant lSI Program were used to define the piping system boundaries. Per Code Case N-716 requirements, piping welds are assigned safety-significance categories, which are then used to determine the examination treatment requirements. High safety-significant (HSS) welds are determined in accordance with the requirements below. Low safety-significant (LSS) welds include all other Class 2,3, or Non-Class welds.

(1)

Class 1 portions of the reactor coolant pressure boundary (RCPB), except as provided in 10 CFR SO.SSa(c)(2)(i) and (c)(2)(ii)

(2)

Applicable portions of the shutdown cooling pressure boundary function. That is, Class 1 and 2 welds of systems or portions of systems needed to utilize the normal shutdown cooling flow path either:

(a) As part of the RCPB from the reactor pressure vessel (RPV) to the second isolation valve (Le., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds; or (b) Other systems or portions of systems from the RPV to the second isolation valve (Le., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds (3)

That portion of the Class 2 feedwater system [greater than 4 inch nominal pipe size (NPS)] of pressurized water reactors (PWRs) from the steam generator to the outer containment isolation valve, (4)

Piping within the break exclusion region (BER) greater than 4 inch NPS for high-energy piping systems as defined by the Owner. Per Code Case N-716, this may include Class 3 or Non-Class piping.

(S)

Any piping segment whose contribution to core damage frequency (CDF) is greater than 1 E-06 [and per NRC feedback on the Grand Gulf and D. C. Cook RIS_B applications 1 E-07 for large early release frequency (LERF)] based upon a plant-specific PSA of pressure boundary failures (e.g., pipe whip, jet impingement, spray, inventory losses). This may include Class 3 or Non-Class piping.

3.2 Failure Potential Assessment Failure potential estimates were generated utilizing industry failure history, plant-specific failure history, and other relevant information. These failure estimates were determined using the guidance provided in NRC approved EPRI TR-1126S7 (Le., the EPRI RI-ISI methodology), with the exception of the deviation discussed below.

Page 11 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table 3.2 summarizes the failure potential assessment by system for each degradation mechanism that was identified as potentially operative.

A deviation to the EPRI RIS_B methodology was implemented in the failure potential assessment for PNP. Table 3-16 of EPRI TR-1 12657 contains the following criteria for assessing the potential for thermal stratification, cycling, and striping (TASCS). Key attributes for horizontal or slightly sloped piping greater than 1 inch NPS include:

1.

The potential exists for low flow in a pipe section connected to a component allowing mixing of hot and cold fluids; or 2.

The potential exists for leakage flow past a valve, including in-leakage, out-leakage and cross-leakage allowing mixing of hot and cold fluids; or 3.

The potential exists for convective heating in dead-ended pipe sections connected to a source of hot fluid; or 4.

The potential exists for two phase (steam/water) flow; or 5.

The potential exists for turbulent penetration into a relatively colder branch pipe connected to header piping containing hot fluid with turbulent flow; AND

• AT> 50°F, AND

• Richardson Number> 4 (this value predicts the potential buoyancy of a stratified flow)

These criteria, based on meeting a high cycle fatigue endurance limit with the AT assumed equal to the greatest potential AT for the transient, will identify locations where stratification is likely to occur, but allows for no assessment of severity. As such, many locations will be identified as subject to TASCS, where no significant potential for thermal fatigue exists. The critical attribute missing from the existing methodology, that would allow consideration of fatigue severity, is a criterion that addresses the potential for fluid cycling. The impact of this additional consideration on the existing TASCS susceptibility criteria is presented below.

Turbulent Penetration TASCS Turbulent penetration is a swirling vertical flow structure in a branch line induced by high velocity flow in the connected piping.

It typically occurs in lines connected to piping containing hot flowing fluid.

In the case of downward sloping lines that then turn horizontal, significant top-to-bottom cyclic ATs can develop in the horizontal sections if the horizontal section is less than about 25 pipe diameters from the reactor coolant piping. Therefore, TASCS is considered for this configuration.

Page 12 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4*20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a{a){3){i)

Table 3.2 summarizes the failure potential assessment by system for each degradation mechanism that was identified as potentially operative.

A deviation to the EPRI RIS_B methodology was implemented in the failure potential assessment for PNP. Table 3-16 of EPRI TR-112657 contains the following criteria for assessing the potential for thermal stratification, cycling, and striping (TASCS). Key attributes for horizontal or slightly sloped piping greater than 1 inch NPS include:

1. The potential exists for low flow in a pipe section connected to a component allowing mixing of hot and cold fluids; or

2. The potential exists for leakage flow past a valve, including in-leakage, out-leakage and cross-leakage allowing mixing of hot and cold fluids; or

3. The potential exists for convective heating in dead-ended pipe sections connected to a source of hot fluid; or

4. The potential exists for two phase (steam/water) flow; or

5. The potential exists for turbulent penetration into a relatively colder branch pipe connected to header piping containing hot fluid with turbulent flow; AND AND

• Richardson Number> 4 (this value predicts the potential buoyancy of a stratified flow)

These criteria, based on meeting a high cycle fatigue endurance limit with the ~ T assumed equal to the greatest potential ~ T for the transient, will identify locations where stratification is likely to occur, but allows for no assessment of severity. As such, many locations will be identified as subject to T ASCS, where no significant potential for thermal fatigue exists. The critical attribute missing from the existing methodology, that would allow consideration of fatigue severity, is a criterion that addresses the potential for fluid cycling. The impact of this additional consideration on the existing T ASCS susceptibility criteria is presented below.

Turbulent Penetration TASeS Turbulent penetration is a swirling vertical flow structure in a branch line induced by high velocity flow in the connected piping. It typically occurs in lines connected to piping containing hot flowing fluid. In the case of downward sloping lines that then turn horizontal, significant top-to-bottom cyclic ~Ts can develop in the horizontal sections if the horizontal section is less than about 25 pipe diameters from the reactor coolant piping. Therefore, T ASCS is considered for this configuration.

Page 12 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

For upward sloping branch lines connected to the hot fluid source that turn horizontal or in horizontal branch lines, natural convective effects combined with effects of turbulence penetration will tend to keep the line filled with hot water.

If there is in-leakage of cold water, a cold stratified layer of water may be formed and significant top-to-bottom ATs may occur in the horizontal portion of the branch line. Interaction with the swirling motion from turbulent penetration may cause a periodic axial motion of the cold layer. Therefore, TASCS is considered for these configurations.

For similar upward sloping branch lines, if there is no potential for in-leakage, this will result in a well-mixed fluid condition where significant top-to-bottom ATs will not occur. Therefore, TASCS is not considered for these no in-leakage configurations. Even in fairly long lines, where some heat loss from the outside of the piping will tend to occur and some fluid stratification may be present, there is no significant potential for cycling as has been observed for the in-leakage case. The effect of TASCS will not be significant under these conditions and can be neglected.

Low flow TASCS In some situations, the transient startup of a system (e.g., shutdown cooling suction piping) creates the potential for fluid stratification as flow is established.

In cases where no cold fluid source exists, the hot flowing fluid will fairly rapidly displace the cold fluid in stagnant lines, while fluid mixing will occur in the piping further removed from the hot source and stratified conditions will exist only briefly as the line fills with hot fluid. As such, since the situation is transient in nature, it can be assumed that the criteria for thermal transients (TT) will govern.

Valve leakage TASCS Sometimes a very small leakage flow of hot water can occur outward past a valve into a line that is relatively colder, creating a significant temperature difference. However, since this is generally a steady-state phenomenon with no potential for cyclic temperature changes, the effect of TASCS is not significant and can be neglected.

Convection Heating TASCS Similarly, there sometimes exists the potential for heat transfer across a valve to an isolated section beyond the valve, resulting in fluid stratification due to natural convection. However, since there is no potential for cyclic temperature changes in this case, the effect of TASCS is not significant and can be neglected.

In summary, these additional considerations for determining the potential for thermal fatigue as a result of the effects of TASCS provide an allowance for considering cycle severity. Consideration of cycle severity was used in previous NRC approved RIS_B program submittals for D. C. Cook (Reference 8), Grand Gulf Nuclear Station (Reference 7), Waterford-3 (Reference 10), and the Vogtle Electric Generating Plant (Reference 11). The methodology used in the PNP RIS_B application for assessing TASCS potential conforms to these updated criteria. Additionally, materials reliability Page 13 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a{a){3){i)

For upward sloping branch lines connected to the hot fluid source that turn horizontal or in horizontal branch lines, natural convective effects combined with effects of turbulence penetration will tend to keep the line filled with hot water. If there is in-leakage of cold water, a cold stratified layer of water may be formed and significant top-to-bottom ~Ts may occur in the horizontal portion of the branch line. Interaction with the swirling motion from turbulent penetration may cause a periodic axial motion of the cold layer. Therefore, TASCS is considered for these configurations.

For similar upward sloping branch lines, if there is no potential for in-leakage, this will result in a well-mixed fluid condition where significant top-to-bottom ~Ts will not occur. Therefore, TASCS is not considered for these no in-leakage configurations. Even in fairly long lines, where some heat loss from the outside of the piping will tend to occur and some fluid stratification may be present, there is no significant potential for cycling as has been observed for the in-leakage case. The effect of T ASCS will not be significant under these conditions and can be neglected.

Low flow T ASCS In some situations, the transient startup of a system (e.g., shutdown cooling suction piping) creates the potential for fluid stratification as flow is established.

In cases where no cold fluid source exists, the hot flowing fluid will fairly rapidly displace the cold fluid in stagnant lines, while fluid mixing will occur in the piping further removed from the hot source and stratified conditions will exist only briefly as the line fills with hot fluid. As such, since the situation is transient in nature, it can be assumed that the criteria for thermal transients (IT) will govern.

Valve leakage TASCS Sometimes a very small leakage flow of hot water can occur outward past a valve into a line that is relatively colder, creating a Significant temperature difference. However, since this is generally a "steady-state" phenomenon with no potential for cyclic temperature changes, the effect of T ASCS is not significant and can be neglected.

Convection Heating TASCS Similarly, there sometimes exists the potential for heat transfer across a valve to an isolated section beyond the valve, resulting in fluid stratification due to natural convection. However, since there is no potential for cyclic temperature changes in this case, the effect of T ASCS is not significant and can be neglected.

In summary, these additional considerations for determining the potential for thermal fatigue as a result of the effects of T ASCS provide an allowance for considering cycle severity. Consideration of cycle severity was used in previous NRC approved RIS_B program submittals for D. C. Cook (Reference 8), Grand Gulf Nuclear Station (Reference 7), Waterford-3 (Reference 10), and the Vogtle Electric Generating Plant (Reference 11). The methodology used in the PNP RIS_B application for assessing T ASCS potential conforms to these updated criteria. Additionally, materials reliability Page 13 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) program (MRP) MRP-146 guidance on the subject of TASCS was also incorporated into the PNP RIS_B application.

3.3 Element and NDE Selection Code Case N-71 6 and lessons learned from the Grand Gulf and DC Cook RIS_B applications provided criteria for identifying the number and location of required examinations. Ten percent of the HSS welds shall be selected for examination as follows:

(1)

Examinations shall be prorated equally among systems to the extent practical, and each system shall individually meet the following requirements:

(a)

A minimum of 25% of the population identified as susceptible to each degradation mechanism and degradation mechanism combination shall be selected.

(b)

If the examinations selected above exceed 10% of the total number of HSS welds, the examinations may be reduced by prorating among each degradation mechanism and degradation mechanism combination, to the extent practical, such that at least 10% of the HSS population is inspected.

(c)

If the examinations selected above are not at least 10% of the HSS weld population, additional welds shall be selected so that the total number selected for examination is at least 10%.

(2)

At least 10% of the RCPB welds shall be selected.

(3)

For the RCPB, at least two-thirds of the examinations shall be located between the inside first isolation valve (IFIV) (i.e., isolation valve closest to the RPV) and the RPV.

(4)

A minimum of 10% of the welds in that portion of the RCPB that lies outside containment (not applicable to PNP) shall be selected.

(5)

A minimum of 10% of the welds within the break exclusion region (BER) shall be selected. This includes main steam and feedwater piping outside containment.

A brief summary of the number of welds and the number selected is provided below, and the results of the selections are presented in Table 3.3. Section 4 of EPRI TR-1 12657 was used as guidance in determining the examination requirements for these locations. Only those RIS_B inspection locations that receive a volumetric examination are included.

Class 1 Welds(

1 H

5

)

Class 2 Welds 2

All Piping Welds 34 Unit Total Selected Total Selected Total Selected 1

626 72 885 19 1523 96 Notes:

(1) Includes all Category B-F and B-J locations except as described in Note 5.

Page l4of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a{a){3){i) program (MRP) MRP-146 guidance on the subject of T ASCS was also incorporated into the PNP RIS_B application.

3.3 Element and NDE Selection Code Case N-716 and lessons learned from the Grand Gulf and DC Cook RIS_B applications provided criteria for identifying the number and location of required examinations. Ten percent of the HSS welds shall be selected for examination as follows:

(1)

Examinations shall be prorated equally among systems to the extent practical, and each system shall individually meet the following requirements:

(a)

A minimum of 250/0 of the population identified as susceptible to each degradation mechanism and degradation mechanism combination shall be selected.

(b)

If the examinations selected above exceed 10% of the total number of HSS welds, the examinations may be reduced by prorating among each degradation mechanism and degradation mechanism combination, to the extent practical, such that at least 10% of the HSS population is inspected.

(c)

If the examinations selected above are not at least 10% of the HSS weld population, additional welds shall be selected so that the total number selected for examination is at least 10 (2)

At least 1 00/0 of the RCPB welds shall be selected.

(3)

For the RCPB, at least two-thirds of the examinations shall be located between the inside first isolation valve (IFIV) (Le., isolation valve closest to the RPV) and the RPV.

(4)

A minimum of 100/0 of the welds in that portion of the RCPB that lies outside containment (not applicable to PNP) shall be selected.

(5)

A minimum of 10

% of the welds within the break exclusion region (BER) shall be selected. This includes main steam and feedwater piping outside containment.

A brief summary of the number of welds and the number selected is provided below, and the results of the selections are presented in Table 3.3. Section 4 of EPRI TR-112657 was used as guidance in determining the examination requirements for these locations. Only those RIS_B inspection locations that receive a volumetric examination are included.

I' Class 1 Welds(1)(5)

Class 2 Welds(2)

J All Piping Welds(3)(4)

Unit Total Selected I'" Total Selected Total Selected 1

626 72 885 19 1523 96 Notes:

(1) Includes all Category 8-F and 8-J locations except as described in Note 5.

Page 14 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

(2) Includes all Category C-F-i and C-F-2 locations. Of the Class 2 piping weld locations, 260 are HSS and the remaining are LSS.

(3) Regardless of safety significance, Class 1, 2, and 3 ASME Section Xl in-scope piping components will continue to be pressure tested as required by the ASME Section XI Program. VT-2 visual examinations are scheduled in accordance with the pressure test program that remains unaffected by the RIS_B Program.

(4) There are 12 non-ASME BER welds in the HSS scope; 5 of these welds were selected for examination. Additional non-ASME piping was also identified as high safety significant and is included in the RIS_B Program although not included in the total weld count.

(5) As described in Section 2.2, Alloy 82/182 welds susceptible to no degradation mechanism or PWSCC only per the RIS_B Program failure potential assessment were removed from the RIS_B population totals in the above table prior to element selection.

3.3.1 Current Examinations PNP is currently using the traditional ASME Section Xl inspection methodology for ISI examination of Class 1 and 2 piping welds per the ASME 2001 Edition with 2003 Addenda.

3.3.2 Successive Examinations If indications are detected during RIS_B ultrasonic examinations, they will be evaluated per IWB-351 4, Standards for Examination Category B-F, Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles, and Examination Category B-J, Pressure Retaining Welds in Piping, (Class 1) or IWC-3514 (Class 2) to determine their acceptability. Any unacceptable flaw will be evaluated per the requirements of ASME Code Section XI, IWB-3600 or IWC-3600, Analytical Evaluation of Flaws, as appropriate. As part of this evaluation, the degradation mechanism that is responsible for the flaw will be determined and accounted for in the evaluation.

If the flaw is acceptable for continued service, successive examinations will be scheduled per Section 6 of Code Case N-71 6. If the flaw is found unacceptable for continued operation, it will be repaired in accordance with IWA-4000, Repair/Replacement Activities, applicable ASME Section XI Code Cases, or NRC approved alternatives. The IWB-3600 analytical evaluation will be submitted to the NRC. Evaluation of indications attributed to PWSCC and successive examinations of PWSCC indications will be performed in accordance with ASME Code Case N-770-1 or a subsequent NRC rule making. Finally, the evaluation will be documented in the corrective action program and the Owner submittals required by Section XI.

3.3.3 Scope Expansion If the nature and type of the flaw is service-induced, then welds subject to the same type of postulated degradation mechanism will be selected and examined per Section 6 of Code Case N-716. The evaluation will include whether other elements in the segment or additional segments are subject to the same degradation mechanisms. Additional examinations will be performed on those elements with the same root cause conditions or degradation mechanisms. The additional examinations will include high safety significant Page 15 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

(2) Includes all Category C-F-1 and C-F-2 locations. Of the Class 2 piping weld locations, 260 are HSS and the remaining are LSS.

(3) Regardless of safety significance, Class 1,2, and 3 ASME Section XI in-scope piping components will continue to be pressure tested as required by the ASME Section XI Program. VT-2 visual examinations are scheduled in accordance with the pressure test program that remains unaffected by the RIS_B Program.

(4) There are 12 non-ASME BER welds in the HSS scope; 5 of these welds were selected for examination. Additional non-ASME piping was also identified as high safety significant and is included in the RIS_B Program although not included in the total weld count.

(5) As described in Section 2.2, Alloy 821182 welds susceptible to no degradation mechanism or PWSCC only per the RIS_B Program failure potential assessment were removed from the RIS_B population totals in the above table prior to element selection.

3.3.1 Current Examinations PNP is currently using the traditional ASME Section XI inspection methodology for lSI examination of Class 1 and 2 piping welds per the ASME 2001 Edition with 2003 Addenda.

3.3.2 Successive Examinations If indications are detected during RIS_B ultrasonic examinations, they will be evaluated per IWB-3514, Standards for Examination Category 8-F, Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles, and Examination Category 8-J, Pressure Retaining Welds in Piping, (Class 1) or IWC-3514 (Class 2) to determine their acceptability. Any unacceptable flaw will be evaluated per the requirements of ASME Code Section XI, IWB-3600 or IWC-3600, Analytical Evaluation of Flaws, as appropriate. As part of this evaluation, the degradation mechanism that is responsible for the flaw will be determined and accounted for in the evaluation. If the flaw is acceptable for continued service, successive examinations will be scheduled per Section 6 of Code Case N-716. If the flaw is found unacceptable for continued operation, it will be repaired in accordance with IWA-4000, Repair/Replacement Activities, applicable ASME Section XI Code Cases, or NRC approved alternatives. The IWB-3600 analytical evaluation will be submitted to the NRC. Evaluation of indications attributed to PWSCC and successive examinations of PWSCC indications will be performed in accordance with ASME Code Case N-770-1 or a subsequent NRC rule making. Finally, the evaluation will be documented in the corrective action program and the Owner submittals required by Section XI.

3.3.3 Scope Expansion If the nature and type of the flaw is service-induced, then welds subject to the same type of postulated degradation mechanism will be selected and examined per Section 6 of Code Case N-716. The evaluation will include whether other elements in the segment or additional segments are subject to the same degradation mechanisms. Additional examinations will be performed on those elements with the same root cause conditions or degradation mechanisms. The additional examinations will include high safety significant Page 15 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

(HSS) elements up to a number equivalent to the number of elements required to be inspected during the current period. If unacceptable flaws or relevant conditions are again found similar to the initial problem, the remaining elements identified as susceptible (includes HSS and low safety significant, (LSS)) will be examined during the current outage. No additional examinations need be performed if there are no additional elements identified as being susceptible to the same degradation mechanism. The need for extensive root cause analysis beyond that required for an IWB-3600, Analytical Evaluation of Flaws, will be dependent on practical considerations (i.e., the practicality of performing additional NDE or removing the flaw for further evaluation during the outage).

Scope expansion for flaws characterized as PWSCC will be conducted in accordance with ASME Code Case N-770-1 or subsequent NRC rule makings.

3.3.4 Program Relief Requests Consistent with previously approved RIS_B submittals, PNP will calculate coverage and use additional examinations or techniques in the same manner it has for traditional Section Xl examinations. Experience has shown this process to be weld-specific (e.g., joint configuration). As such, the effect on risk, if any, will not be known until the examinations are performed. Relief requests for those cases where greater than 90% coverage is not obtained will be submitted per the requirements of 10 CFR 50.55a(g)(5)(iv).

As discussed in EPRI TR-1 12657, accessibility is an important consideration in the element selection process of a RI-ISI application. As such, for the PNP N-71 6 application, locations will generally be selected for examination where the desired coverage is achievable. This is typically accomplished by utilizing previous inspection history, plant access considerations, and knowledgeable plant personnel. However, some limitations will not be known until the examination is performed since some locations will be examined for the first time. In addition, other considerations may take precedence and dictate the selection of locations where greater than 90% examination coverage is physically impossible. This is especially true for element selections where a degradation mechanism may be operative (e.g., risk categories 1, 2, 3 and 5 of EPRI TR-112657). For these locations, elements are generally selected for examination on the basis of predicted degradation severity. For example, in the emergency core cooling system (ECCS) injection lines of PWRs, the piping section immediately upstream of the first isolation check valve is considered susceptible to intergranular stress corrosion cracking (IGSCC), assuming a sufficiently high temperature and oxygenated water supply. The piping element (pipe-to-valve weld) located nearest the heat source will be subjected to the highest temperature (conduction heating). As such, this location will generally be selected for examination since it is considered more susceptible than locations further removed from the heat source, even though a pipe-to-valve weld is inherently more difficult to examine and obtain full coverage than most other configurations (e.g., pipe-to-elbow weld). In this example, less than 90%

coverage of this location will yield far more valuable information than 100%

coverage of a less susceptible location.

Page 16 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.S5a{a){3){i)

(HSS) elements up to a number equivalent to the number of elements required to be inspected during the current period. If unacceptable flaws or relevant conditions are again found similar to the initial problem, the remaining elements identified as susceptible (includes HSS and low safety significant, (LSS>> will be examined during the current outage. No additional examinations need be performed if there are no additional elements identified as being susceptible to the same degradation mechanism. The need for extensive root cause analysis beyond that required for an IWB-3600, Analytical Evaluation of Flaws, will be dependent on practical considerations (Le., the practicality of performing additional NDE or removing the flaw for further evaluation during the outage).

Scope expansion for flaws characterized as PWSCC will be conducted in accordance with ASME Code Case N-770-1 or subsequent NRC rule makings.

3.3.4 Program Relief Requests Consistent with previously approved RIS_B submittals, PNP will calculate coverage and use additional examinations or techniques in the same manner it has for traditional Section XI examinations. Experience has shown this process to be weld-specific (e.g., joint configuration). As such, the effect on risk, if any, will not be known until the examinations are performed. Relief requests for those cases where greater than 900/0 coverage is not obtained will be submitted per the requirements of 10 CFR 50.55a(g)(5)(iv).

As discussed in EPRI TR-112657, accessibility is an important consideration in the element selection process of a RI-ISI application. As such, for the PNP N-716 application, locations will generally be selected for examination where the desired coverage is achievable. This is typically accomplished by utilizing previous inspection history, plant access considerations, and knowledgeable plant personnel. However, some limitations will not be known until the examination is performed since some locations will be examined for the first time. In addition, other considerations may take precedence and dictate the selection of locations where greater than 90% examination coverage is physically impossible. This is especially true for element selections where a degradation mechanism may be operative (e.g., risk categories 1,2,3 and 5 of EPRI TR-112657). For these locations, elements are generally selected for examination on the basis of predicted degradation severity. For example, in the emergency core COOling system (ECCS) injection lines of PWRs, the piping section immediately upstream of the first isolation check valve is considered susceptible to intergranular stress corrosion cracking (IGSCC), assuming a sufficiently high temperature and oxygenated water supply. The piping element (pipe-to-valve weld) located nearest the heat source will be subjected to the highest temperature (conduction heating). As such, this location will generally be selected for examination since it is considered more susceptible than locations further removed from the heat source, even though a pipe-to-valve weld is inherently more difficult to examine and obtain full coverage than most other configurations (e.g., pipe-to-elbow weld). In this example, less than 90%

coverage of this location will yield far more valuable information than 100%

coverage of a less susceptible location.

Page 16 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

For locations with no identified degradation mechanisms (i.e., similar to risk category 4 of EPRI TR-112657), a greater degree of flexibility exists in choosing inspection locations. As such, if at the time of examination an N-716 element selection is found to be obstructed, a more suitable location may be substituted instead.

Therefore, ENO will review each instance of limited coverage and take the appropriate steps (e.g., relief requests) consistent with its impact on the basis of the N-716 application.

No Palisades relief requests are being withdrawn due to the RIS_B application.

3.4 Risk Impact Assessment The RIS_B Program development has been conducted in accordance with RG 1.174 and the requirements of Code Case N-71 6, and the risk from implementation of this program is expected to remain neutral or decrease when compared to that estimated from current requirements.

This evaluation categorized welds as high safety significant or low safety significant in accordance with Code Case N-71 6, and then determined what inspection changes were proposed for each system. The changes included changing the number and location of inspections, and in many cases improving the effectiveness of the inspection to account for the findings of the RIS_B degradation mechanism assessment. For example, examinations of locations subject to thermal fatigue will be conducted on an expanded volume and will be focused to enhance the probability of detection (POD) during the inspection process.

3.4.1 Quantitative Analysis Code Case N-71 6 has adopted the NRC approved EPRI TR-1 12657 process for risk impact analyses, whereby limits are imposed to ensure that the change-in-risk of implementing the RIS_B Program meets the requirements of RG 1.174 and RG 1.178. Section 3.7.2 of EPRI TR-112657 requires that the cumulative change in CDF and LERF be less than 1 E-07 and 1 E-08 per year per system, respectively.

For LSS welds, conditional core damage probability (CCDP)/conditional large early release probability (CLERP) values of 1 E-4/1 E-5 were conservatively used except for FWS and MSS welds where actual values were available. The rationale for using the 1 E-4/1 E-5 values is that the change-in-risk evaluation process of Code Case N-71 6 is similar to that of the EPRI RI-ISI methodology.

As such, the goal is to determine CCDPs/CLERPs threshold values. For example, the threshold values between high and medium consequence categories is 1 E-4 (CCDP)/1 E-5 (CLERP) and between medium and low consequence categories are 1 E-6 (CCDP)/1 E-7 (CLERP) from the EPRI RI-ISI Risk Matrix. Using these threshold values streamlines the change-in-risk evaluation as well as stabilizes the update process. For example, if a CCDP changes from 1 E-5 to 3E-5 due to an update, it will remain below the 1 E-4 threshold value; the change-in-risk evaluation would not require updating.

Page 17 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.558 RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

For locations with no identified degradation mechanisms (Le., similar to risk category 4 of EPRI TR-112657), a greater degree of flexibility exists in choosing inspection locations. As such, if at the time of examination an N-716 element selection is found to be obstructed, a more suitable location may be substituted instead.

Therefore, ENO will review each instance of limited coverage and take the appropriate steps (e.g., relief requests) consistent with its impact on the basis of the N-716 application.

No Palisades relief requests are being withdrawn due to the RIS_B application.

3.4 Risk Impact Assessment The RIS_B Program development has been conducted in accordance with RG 1.174 and the requirements of Code Case N-716, and the risk from implementation of this program is expected to remain neutral or decrease when compared to that estimated from current requirements.

This evaluation categorized welds as high safety significant or low safety significant in accordance with Code Case N-716, and then determined what inspection changes were proposed for each system. The changes included changing the number and location of inspections, and in many cases improving the effectiveness of the inspection to account for the findings of the RIS_B degradation mechanism assessment. For example, examinations of locations subject to thermal fatigue will be conducted on an expanded volume and will be focused to enhance the probability of detection (POD) during the inspection process.

3.4.1 Quantitative Analysis Code Case N-716 has adopted the NRC approved EPRI TR-112657 process for risk impact analyses, whereby limits are imposed to ensure that the change-in-risk of implementing the RIS_B Program meets the requirements of RG 1.174 and RG 1.17B. Section 3.7.2 of EPRI TR-112657 requires that the cumulative change in CDF and LERF be less than 1 E-07 and 1 E-OB per year per system, respectively.

For LSS welds, conditional core damage probability (CCDP)/conditional large early release probability (CLERP) values of 1 E-4/1 E-5 were conservatively used except for FWS and MSS welds where actual values were available. The rationale for using the 1 E-4/1 E-5 values is that the change-in-risk evaluation process of Code Case N-716 is similar to that of the EPRI RI-ISI methodology.

As such, the goal is to determine CCDPs/CLERPs threshold values. For example, the threshold values between high and medium consequence categories is 1 E-4 (CCDP)/1 E-5 (CLERP) and between medium and low consequence categories are 1 E-6 (CCDP)/1 E-7 (CLERP) from the EPRI RI-ISI Risk Matrix. Using these threshold values streamlines the change-in-risk evaluation as well as stabilizes the update process. For example, if a CCDP changes from 1 E-5 to 3E-5 due to an update, it will remain below the 1 E-4 threshold value; the change-in-risk evaluation would not require updating.

Page 17 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

The updated internal flooding PRA was also reviewed to ensure that there is no LSS Class 2 piping with a CCDP/CLERP greater than 1 E-4/1 E-5.

With respect to assigning failure potentials for LSS piping, the criteria are defined in Table 3 of Code Case N-716. That is, those locations identified as susceptible to FAC are assigned a high failure potential. Those locations susceptible to thermal fatigue, erosion-cavitation, corrosion, or stress corrosion cracking are assigned a medium failure potential, unless they have an identified potential for water hammer loads. In such cases, they will be assigned a high failure potential. Finally, those locations that are identified as not susceptible to degradation are assigned a low failure potential.

In order to streamline the risk impact assessment, a review was conducted that verified that the LSS piping was not susceptible to water hammer. LSS piping may be susceptible to FAC; however, the examination for FAC is performed per the FAC program. This review was conducted similar to that done for a traditional RI-ISI application. Thus, the high failure potential category is not applicable to LSS piping. In lieu of conducting a formal degradation mechanism evaluation for all LSS piping (e.g. to determine if thermal fatigue is applicable),

these locations were conservatively assigned to the medium failure potential (assume medium in Table 3.4) for use in the change-in-risk assessment.

Experience with previous industry RIS_B applications shows this to be conservative.

PNP has conducted a risk impact analysis per the requirements of Section 5 of Code Case N-71 6 that is consistent with the simplified risk quantification method described in Section 3.7 of EPRI TR-112657. The analysis estimates the net change-in-risk due to the positive and negative influences of adding and removing locations from the inspection program.

The CCDP and CLERP values used to assess risk impact were estimated based on pipe break location. Based on these estimated values, a corresponding consequence rank was assigned per the requirements of EPRI TR-112657 and upper bound threshold values were used as provided in the table below. Consistent with the EPRI methodology, the upper bound for all break locations that fall within the high consequence rank range was based on the highest CCDP value obtained (e.g., large LOCA CCDP bounds the medium and small LOCA CCDP5).

Also, as described in Section 2.2, Alloy 82/182 welds susceptible to no degradation mechanism or PWSCC only per the RIS_B Program failure potential assessment were removed from the RIS_B population prior to element selection and risk impact assessment.

Page 18 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

The updated internal flooding PRA was also reviewed to ensure that there is no LSS Class 2 piping with a CCDP/CLERP greater than 1 E-4/1 E-5.

With respect to assigning failure potentials for LSS piping, the criteria are defined in Table 3 of Code Case N-716. That is, those locations identified as susceptible to FAC are assigned a high failure potential. Those locations susceptible to thermal fatigue, erosion-cavitation, corrosion, or stress corrosion cracking are assigned a medium failure potential, unless they have an identified potential for water hammer loads. In such cases, they will be assigned a high failure potential. Finally, those locations that are identified as not susceptible to degradation are assigned a low failure potential.

In order to streamline the risk impact assessment, a review was conducted that verified that the LSS piping was not susceptible to water hammer. LSS piping may be susceptible to FAC; however, the examination for FAC is performed per the FAC program. This review was conducted similar to that done for a traditional RI-ISI application. Thus, the high failure potential category is not applicable to LSS piping. In lieu of conducting a formal degradation mechanism evaluation for all LSS piping (e.g. to determine if thermal fatigue is applicable),

these locations were conservatively assigned to the medium failure potential

("assume medium" in Table 3.4) for use in the change-in-risk assessment.

Experience with previous industry RIS_B applications shows this to be conservative.

PNP has conducted a risk impact analysis per the requirements of Section 5 of Code Case N-716 that is consistent with the "simplified risk quantification method" described in Section 3.7 of EPRI TR-112657. The analysis estimates the net change-in-risk due to the positive and negative influences of adding and removing locations from the inspection program.

The CCDP and CLERP values used to assess risk impact were estimated based on pipe break location. Based on these estimated values, a corresponding consequence rank was assigned per the requirements of EPRI TR-112657 and upper bound threshold values were used as provided in the table below. Consistent with the EPRI methodology, the upper bound for all break locations that fall within the high consequence rank range was based on the highest CCDP value obtained (e.g., large LOCA CCDP bounds the medium and small LOCA CCDPs).

Also, as described in Section 2.2, Alloy 821182 welds susceptible to no degradation mechanism or PWSCC only per the RIS_B Program failure potential assessment were removed from the RIS_B population prior to element selection and risk impact assessment.

Page 18 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Break I

Estimated Consequence Upper I Lower Bound Location I CCDP CLERP Rank CCDP CLERP Description of Affected Piping LOCA 9E-03 9E-04 The highest CCDP for Large LOCA HIGH (U) 9E-03 (U) 9E-04 Unisolable RCPB piping of all sizes (IE_LBLOCA) was used with 0.1 LERP (L) 1 E-04 (L) 1 E-05 ILOCA 3E-05 3E-06 Calculated based on LOCA (U) 1 E-04 (U) 1 E-05 Piping between 1st and 2nd normally (IE_LBLOCA) CCDP and CLERP times MEDIUM (L) 1 E-06 (L) 1 E-07 open isolation valve inside containment valve fail to close probability 3E-3 (letdown and charging)

PLOCA 9E-06 9E-07 Calculated based on LOCA (U) 1 E-04 (U) 1 E-05 Piping between 1st and 2nd normally (IE_LBLOCA) CCDP and CLERP times MEDIUM (L) 1 E-06 (L) 1 E-07 closed isolation valve inside valve internal rupture 1 E-3 containment (aux spray, ECCS, drains)

PPLOCA

<1E-06

<1E-07 Calculated based on LOCA (U) 1 E06 (U) 1 E07 (IE_LBLOCA) CCDP and CLERP times LOW (L) 1 E-06 (L) 1 E-07 SDC Class 2 piping 2 valves in series internal rupture <1 E-4 MSLB-l 1 E-05 1 E-06 Main Steam and Feedwater line breaks MEDIUM (V) 1 E-04 (V) 1 E-05 MS and FW piping inside containment inside containment CCDP and 0.1 LERP (M) 1 E-06 (M) 1 E-07 MSLB-O 1 E-02 1 E-03 Main Steam and Feedwater line breaks HIGH (U) 1 E-02 (U) 1 E-03 MS and FW piping outside containment outside containment CCDP and 0.1 (L) 1 E-04 (L) 1 E-05 LERP Class2 1 E-04 I

1 E-05 LSS I

(U) 1 E-04 (U) 1 E-05 All Class 2 system piping designated as Estimated based on upper bound for MEDIUM (L) 1 E-06 (L) 1 E-07 low safety significant except for piping Medium Consequence MS and FW piping 1.

The PRA does not explicitly model potential and isolable LOCA events, because such events are subsumed by the LOCA initiators in the PRA. That is, the frequency of a LOCA in this limited piping downstream of the first RCPB isolation valve times the probability that the valve fails is a small contributor to the total LOCA frequency.

The N-716 methodology must evaluate these segments individually; thus, it is necessary to estimate their contribution.

This is estimated by taking the LOCA CCDP and multiplying it by the valve failure probability.

PLOCA is identified and used in the quantification of both ILOCA (isolable LOCA) and PLOCA The likelihood of pressure boundary failure (PBF) is determined by the presence of different degradation mechanisms and the rank is based on the relative failure probability. The basic likelihood of PBF for a piping location with no degradation mechanism present is given as x0 and is expected to have a value less than 1 E-08. Piping locations identified as medium failure potential have a likelihood of 20x

0. These PBF likelihoods are consistent with References 9 and 14 of EPRI TR-112657. In addition, the analysis was performed both with and without taking credit for enhanced inspection effectiveness due to an increased POD from application of the RIS_B approach.

Table 3.4 presents a summary of the RIS_B Program versus the first ISI interval (1977 Edition/i 978 Addendum of ASME Section Xl) program requirements on a per system basis. The presence of FAC was adjusted for in the quantitative analysis by excluding its impact on the failure potential rank.

The exclusion of the impact of FAC on the failure potential rank, and therefore in the determination of the change-in-risk, was performed because FAC is a Page 19 of 30 Break Location LOCA ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Estimated Consequence Upper I Lower Bound Description of Affected Piping CCDP I

CLERP Rank CCDP CLERP 9E-03 I

9E-04 (U) 9E-03 (U) 9E-04 The highest CCDP for Large LOCA HIGH (L) 1E-04 (L) 1E-05 Unisolable RCPB piping of all sizes (IE LBLOCA) was used with 0.1 LERP I LOCA 3E-05 I

3E-06 Piping between 1 st and 2nd normally Calculated based on LOCA MEDIUM (U) 1E-04 (U) 1E-05 open isolation valve inside containment (IE_LBLOCA) CCDP and CLERP times (L) 1E-06 (L) 1E-07 (letdown and charging) valve fail to close probability -3E-3 PLOCA 9E-06 I

9E-07 Piping between 1 st and 2nd normally Calculated based on LOCA MEDIUM (U) 1E-04 (U) 1E-05 closed isolation valve inside (IE_LBLOCA) CCDP and CLERP times (L) 1E-06 (L) 1E-07 containment (aux spray, ECCS, drains) valve internal rupture -1 E-3 PPLOCA

<1E-06 I

<1E-07 Calculated based on LOCA LOW (U) 1E-06 (U) 1E-07 SDC Class 2 piping (IE_LBLOCA) CCDP and CLERP times (L) 1E-06 (L) 1E-07 2 valves in series internal rupture <1 E-4 MSLB-I 1E-05 I

1E-06 (V) 1E-04 (V) 1E-05 Main Steam and Feedwater line breaks MEDIUM (M) 1E-06 (M) 1E-07 MS and FW piping inside containment inside containment CCDP and 0.1 LERP MSLB-O 1E-02 I

1E-03 Main Steam and Feedwater line breaks HIGH (U) 1E-02 (U) 1E-03 MS and FW piping outside containment outside containment CCDP and 0.1 (L) 1E-04 (L) 1E-05 LERP Class 2 1E-04 I 1E-05 All Class 2 system piping designated as LSS MEDIUM (U) 1E-04 (U) 1E-05 low safety significant except for piping Estimated based on upper bound for (L) 1E-06 (L) 1E-07 MS and FW piping Medium Consequence

1. The PRA does not explicitly model potential and isolable LOCA events, because such events are subsumed by the LOCA initiators in the PRA. That is, the frequency of a LOCA in this limited piping downstream of the first RCPS isolation valve times the probability that the valve fails is a small contributor to the total LOCA frequency. The N-716 methodology must evaluate these segments individually; thus, it is necessary to estimate their contribution. This is estimated by taking the LOCA CCOP and multiplying it by the valve failure probability. PLOCA is identified and used in the quantification of both ILOCA (isolable LOCA) and PLOCA The likelihood of pressure boundary failure (PBF) is determined by the presence of different degradation mechanisms and the rank is based on the relative failure probability. The basic likelihood of PBF for a piping location with no degradation mechanism present is given as Xo and is expected to have a value less than 1 E-OB. Piping locations identified as medium failure potential have a likelihood of 20Xo. These PBF likelihoods are consistent with References 9 and 14 of EPRI TR-112657. In addition, the analysis was performed both with and without taking credit for enhanced inspection effectiveness due to an increased POD from application of the RIS_B approach.

Table 3.4 presents a summary of the RIS_B Program versus the first lSI interval (1977 Edition/197B Addendum of ASME Section XI) program requirements on a "per system" basis. The presence of FAC was adjusted for in the quantitative analysis by excluding its impact on the failure potential rank.

The exclusion of the impact of FAC on the failure potential rank, and therefore in the determination of the change-in-risk, was performed because FAC is a Page 19 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) damage mechanism managed by a separate, independent plant augmented inspection program. The RIS_B Program credits and relies upon this plant augmented inspection program to manage this damage mechanism. The plant FAC program will continue to determine where and when examinations shall be performed. Hence, since the number of FAC examination locations remains the same before and after (the implementation of the RIS_B program) and no delta exists, there is no need to include the impact of FAC in the performance of the risk impact analysis.

As indicated in the following table, this evaluation has demonstrated that unacceptable risk impacts will not occur from implementation of the RIS_B Program, and that the acceptance criteria of RG 1.174 and Code Case N-716 are satisfied.

S With POD Credit Without POD Credit ys em Delta CDF Delta LERF Delta CDF Delta LERF CVCs

- Chemical Volume & Control System

-4.86E-09

-4.86E-10

-2.70E-09

-2.70E-10 ESS

- Engineered Safeguards System 1.06E-1 0 1.06E-1 1 1.06E-1 0 1.06E-1 1 FWS Feedwater System 1.97E-09 1.97E-1 0 1.97E-09 1.97E-1 0 MSS

- Main Steam System

-4.50E-1 0

-4.50E-1 1

-4.50E-1 0

-4.50E-1 1 PCS

- Primary Coolant System

-1.78E-08

-1.78E-09

-7.70E-09

-7.70E-1 0 RWS

- Radwaste System 0.00E+0O 0.00E+0O 0.00E+00 O.OOE+00 SWS

- Service Water System 3.OOE-1 1 3.OOE-1 2 3.OOE-1 1 3.OOE-1 2 VAS

- Vent & Air Conditioning System 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Total

-2.1OE-08

-2.1OE-09

-8.75E-09

-8.75E-10 Note:

(1)

The risk reduction associated with Class 3 and non-ASME piping is not shown in the above table. However, 12 welds associated with non-ASME BER piping in FWS (8) and MSS (4) are included in the above table. Inspections of this piping will lead to a risk reduction.

As shown in Table 3.4, new RIS_B locations were selected such that the RIS_B selections exceed the Section Xl selections for certain categories (Delta column has a positive number). To show that the use of a conservative upper bound CCDP/CLERP does not result in an optimistic calculation with regard to meeting the acceptance criteria, a conservative sensitivity was conducted where the RIS_B selections were set equal to the Section Xl selections (Delta changed from positive number to zero). The acceptance criteria are met when the number of RIS_B selections is not allowed to exceed Section Xl.

3.4.2 Defense-in-Depth The intent of the inspections mandated by 10 CFR 50.55a for piping welds is to identify conditions such as flaws or indications that may be precursors to leaks or ruptures in a systems pressure boundary. Currently, the process for selecting inspection locations is based upon terminal end locations, structural discontinuities, and stress analysis results. As depicted in ASME White Paper 92-01-01, Rev. 1, Evaluation of Inservice Inspection Requirements for Class 1, Page 20 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) damage mechanism managed by a separate, independent plant augmented inspection program. The RIS_B Program credits and relies upon this plant augmented inspection program to manage this damage mechanism. The plant FAC program will continue to determine where and when examinations shall be performed. Hence, since the number of FAC examination locations remains the same "before" and "after" (the implementation of the RIS_B program) and no delta exists, there is no need to include the impact of FAC in the performance of the risk impact analysis.

As indicated in the following table, this evaluation has demonstrated that unacceptable risk impacts will not occur from implementation of the RIS_B Program, and that the acceptance criteria of RG 1.174 and Code Case N-716 are satisfied.

System With POD Credit Without POD Credit Delta CDF Delta LERF Delta CDF Delta LERF CVCs - Chemical Volume & Control System

-4.B6E-09

-4.B6E-10

-2.70E-09

-2.70E-10 ESS - Engineered Safeguards System 1.06E-10 1.06E-11 1.06E-10 1.06E-11 FWS - Feedwater System 1.97E-09 1.97E-10 1.97E-09 1.97E-10 MSS - Main Steam System

-4.50E-10

-4.50E-11

-4.50E-10

-4.50E-11 PCS - Primary Coolant System

-1.7BE-OB

-1.7BE-09

-7.70E-09

-7.70E-10 RWS - Radwaste System O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO SWS - Service Water System 3.00E-11 3.00E-12 3.00E-11 3.00E-12 VAS - Vent & Air Conditioning System O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO Total

-2.10E-OS

-2.10E-09

-S.75E-09

-S.75E-10 Note:

(1)

The risk reduction associated with Class 3 and non-ASME piping is not shown in the above table. However, 12 welds associated with non-ASME SER piping in FWS (B) and MSS (4) are included in the above table. Inspections of this piping will lead to a risk reduction.

As shown in Table 3.4, new RIS_B locations were selected such that the RIS_B selections exceed the Section XI selections for certain categories (Delta column has a positive number). To show that the use of a conservative upper bound CCDP/CLERP does not result in an optimistic calculation with regard to meeting the acceptance criteria, a conservative sensitivity was conducted where the RIS_B selections were set equal to the Section XI selections (Delta changed from positive number to zero). The acceptance criteria are met when the number of RIS_B selections is not allowed to exceed Section XI.

3.4.2 Defense-in-Depth The intent of the inspections mandated by 10 CFR 50.55a for piping welds is to identify conditions such as flaws or indications that may be precursors to leaks or ruptures in a system's pressure boundary. Currently, the process for selecting inspection locations is based upon terminal end locations, structural discontinuities, and stress analysis results. As depicted in ASME White Paper 92-01-01, Rev. 1, Evaluation of InseNice Inspection Requirements for Class 1, Page 20 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Category B-J Pressure Retaining Welds, this methodology has been ineffective in identifying leaks or failures. EPRI TR-112657 and Code Case N-716 provide a more robust selection process founded on actual service experience with nuclear plant piping failure data.

This process has two key independent ingredients; that is, a determination of each locations susceptibility to degradation and secondly, an independent assessment of the consequence of the piping failure. These two ingredients assure defense-in-depth is maintained. First, by evaluating a locations susceptibility to degradation, the likelihood of finding flaws or indications that may be precursors to leak or ruptures is increased. Secondly, a generic assessment of high-consequence sites has been determined by Code Case N-716, supplemented by plant-specific evaluations, thereby requiring a minimum threshold of inspection for important piping whose failure would result in a LOCA or BER break. Finally, Code Case N-71 6 requires that any piping on a plant-specific basis that has a contribution to CDF of greater than 1 E-06 (or 1 E-07 for LERF) be included in the scope of the application. PNP identified non-ASME piping as HSS.

All locations within the Class 1, 2, and 3 pressure boundaries will continue to be pressure tested in accordance with the Code, regardless of its safety significance.

3.5 Implementation Program Upon approval of the RIS_B Program, procedures that comply with the guidelines described in Code Case N-71 6 will be prepared to implement and monitor the program. The new program will be implemented during the third period of the fourth interval. No changes to the Operating License, Technical Specifications or Updated Final Safety Analysis Report are necessary for program implementation.

The applicable aspects of the ASME Code not affected by this change will be retained, such as inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements. Existing ASME Section Xl program implementing procedures will be retained and modified to address the RIS_B process, as appropriate.

3.6 Feedback (Monitoring)

The RIS_B Program is a living program that is required to be monitored continuously for changes that could impact the basis for which welds are selected for examination.

Monitoring encompasses numerous facets, including the review of changes to the plant configuration, changes to operations that could affect the degradation assessment, a review of NDE results, a review of site failure information from the corrective action program, and a review of industry failure information from industry operating experience (OE). Also included is a review of PRA changes for their impact on the RIS_B program. These reviews provide a feedback loop such that new relevant information is obtained that will ensure that the appropriate identification of HSS piping locations selected for examination is maintained. As a minimum, this review will be Page 21 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Category 8-J Pressure Retaining Welds, this methodology has been ineffective in identifying leaks or failures. EPRI TR-112657 and Code Case N-716 provide a more robust selection process founded on actual service experience with nuclear plant piping failure data.

This process has two key independent ingredients; that is, a determination of each location's susceptibility to degradation and secondly, an independent assessment of the consequence of the piping failure. These two ingredients assure defense-in-depth is maintained. First, by evaluating a location's susceptibility to degradation, the likelihood of finding flaws or indications that may be precursors to leak or ruptures is increased. Secondly, a generic assessment of high-consequence sites has been determined by Code Case N-716, supplemented by plant-specific evaluations, thereby requiring a minimum threshold of inspection for important piping whose failure would result in a LOCA or BER break. Finally, Code Case N-716 requires that any piping on a plant-specific basis that has a contribution to CDF of greater than 1 E-06 (or 1 E-07 for LERF) be included in the scope of the application. PNP identified non-ASME piping as HSS.

All locations within the Class 1, 2, and 3 pressure boundaries will continue to be pressure tested in accordance with the Code, regardless of its safety significance.

3.5 Implementation Program Upon approval of the RIS_B Program, procedures that comply with the guidelines described in Code Case N-716 will be prepared to implement and monitor the program. The new program will be implemented during the third period of the fourth interval. No changes to the Operating License, Technical Specifications or Updated Final Safety Analysis Report are necessary for program implementation.

The applicable aspects of the ASME Code not affected by this change will be retained, such as inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements. Existing ASME Section XI program implementing procedures will be retained and modified to address the RIS_B process, as appropriate.

3.6 Feedback (Monitoring)

The RIS_B Program is a living program that is required to be monitored continuously for changes that could impact the basis for which welds are selected for examination.

Monitoring encompasses numerous facets, including the review of changes to the plant configuration, changes to operations that could affect the degradation assessment, a review of NDE results, a review of site failure information from the corrective action program, and a review of industry failure information from industry operating experience (OE). Also included is a review of PRA changes for their impact on the RIS_B program. These reviews provide a feedback loop such that new relevant information is obtained that will ensure that the appropriate identification of HSS piping locations selected for examination is maintained. As a minimum, this review will be Page 21 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) conducted on an ASME period basis. In addition, more frequent adjustment may be required as directed by NRC Bulletin or Generic Letter requirements, or by industry and plant-specific feedback.

If an adverse condition, such as an unacceptable flaw is detected during examinations, the adverse condition will be addressed by the corrective action program and procedures. The following are appropriate actions to be taken:

A.

Identify (Examination results conclude there is an unacceptable flaw).

B.

Characterize (Determine if regulatory reporting is required and assess if an immediate safety or operation impact exists).

C.

Evaluate (Determine the cause and extent of the condition identified and develop a corrective action plan or plans).

D.

Decide (Make a decision to implement the corrective action plan).

E.

Implement (Complete the work necessary to correct the problem and prevent recurrence).

F.

Monitor (Through the audit process ensure that the RIS_B program has been updated based on the completed corrective action).

G.

Trend (Identify conditions that are significant based on accumulation of similar issues).

For preservice examinations, PNP will follow the rules contained in Section 3.0 of N-716. Welds classified HSS require a preservice inspection. The examination volumes, techniques, and procedures shall be in accordance with Table 1 of N-716.

Welds classified as LSS do not require preservice inspection.

4.

PROPOSED ISI PLAN CHANGE PNP intends to start implementing the RIS_B Program during the plants third period of the current (fourth) inspection interval. By the end of second period of the fourth lSl interval, 56% of the piping weld examinations required by ASME Section Xl had been completed for Examination Categories B-F, B-J, C F-i and C-F-2, with the remaining 44% of the examinations to be completed during the two refueling outages in the third period (1 R23 and 1 R24). The examinations in the first outage of the third period (1 R23) were completed under the ASME Section XI requirements while the examinations in the second outage of the third period (1 R24) will be completed per the RIS_B process. Examinations will be performed such that the period percentage requirements of ASME Section XI are met.

As discussed in Section 2.2, implementation of the RIS_B program will not alter any PWSCC examination requirements for the Alloy 82/182 examinations.

A comparison between the RIS_B Program and the 2001 Edition through the 2003 Addendum of Section XI program requirements for fourth interval in-scope piping is provided in Table 4. In addition, non-ASME piping was identified as high safety significant and is included in the RIS_B Program. Ten percent of the welds will be inspected in accordance with N-7i 6.

Page 22 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) conducted on an ASME period basis. In addition, more frequent adjustment may be required as directed by NRC Bulletin or Generic Letter requirements, or by industry and plant-specific feedback.

If an adverse condition, such as an unacceptable flaw is detected during examinations, the adverse condition will be addressed by the corrective action program and procedures. The following are appropriate actions to be taken:

A.

Identify (Examination results conclude there is an unacceptable flaw).

B.

Characterize (Determine if regulatory reporting is required and assess if an immediate safety or operation impact exists).

C.

Evaluate (Determine the cause and extent of the condition identified and develop a corrective action plan or plans).

D.

Decide (Make a decision to implement the corrective action plan).

E.

Implement (Complete the work necessary to correct the problem and prevent recurrence).

F.

Monitor (Through the audit process ensure that the RIS_B program has been updated based on the completed corrective action).

G.

Trend (Identify conditions that are significant based on accumulation of similar issues).

For preservice examinations, PNP will follow the rules contained in Section 3.0 of N-716. Welds classified HSS require a preservice inspection. The examination volumes, techniques, and procedures shall be in accordance with Table 1 of N-716.

Welds classified as LSS do not require preservice inspection.

4.

PROPOSED lSI PLAN CHANGE PNP intends to start implementing the RIS_B Program during the plant's third period of the current (fourth) inspection interval. By the end of second period of the fourth lSI interval, 560/0 of the piping weld examinations required by ASME Section XI had been completed for Examination Categories B-F, B-J, C F-1 and C-F-2, with the remaining 440/0 of the examinations to be completed during the two refueling outages in the third period (1 R23 and 1 R24). The examinations in the first outage of the third period (1 R23) were completed under the ASME Section XI requirements while the examinations in the second outage of the third period (1 R24) will be completed per the RIS_B process. Examinations will be performed such that the period percentage requirements of ASME Section XI are met.

As discussed in Section 2.2, implementation of the RIS_B program will not alter any PWSCC examination requirements for the Alloy 821182 examinations.

A comparison between the RIS_B Program and the 2001 Edition through the 2003 Addendum of Section XI program requirements for fourth interval in-scope piping is provided in Table 4. In addition, non-ASME piping was identified as high safety significant and is included in the RIS_B Program. Ten percent of the welds will be inspected in accordance with N-716.

Page 22 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

The degradation mechanism identified for the FPS piping is pitting, and the examination will be performed in accordance with the owners existing localized corrosion program. The examination volume shall include base metal, welds, and weld HAZ in the affected regions of carbon and low-alloy steel, and the welds and weld HAZ of austenitic steel. Examinations shall verify the minimum wall thickness required. The examination method and examination region shall be sufficient to characterize the extent of the element degradation.

5.

REFERENCES/DOCUMENTATION 1.

EPRI Report 1006937, Extension of EPRI Risk Informed IS! Methodology to Break Exclusion Region Programs.

2.

EPRI TR-1 12657, Revised Risk-Informed Inservice Inspection Evaluation Procedure, Rev. B-A.

3.

ASME Code Case N-71 6, Alternative Piping Classification and Examination Requirements, Section Xl Division 1.

4.

Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis.

5.

Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping.

6.

Regulatory Guide 1.200, Rev 2 An Approach For Determining The Technical Adequacy Of Probabillstic Risk Assessment Results For Risk-Informed Activities.

7.

USNRC Safety Evaluation for Grand Gulf Nuclear Station Unit 1, Request for Alternative GG-ISl-002-lmplement Risk-Informed ISl based on ASME Code Case N-71 6, dated September 21, 2007. ADAMS Accession No. ML072430005 8.

USNRC Safety Evaluation for DC Cook Nuclear Plant, Units 1 and 2, Risk-Informed Safety-Based ISI program for Class 1 and 2 Piping Welds, dated September 28, 2007. See ADAMS Accession No. ML072620553.

9.

EPRI Report 1021467 Nondestructive Evaluation: Probabillstic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs.

10. Waterford-3 Safety Evaluation See ADAMS Accession No. ML080980120.

11. Vogtle Electric Generating Plant Safety Evaluation

- See ADAMS Accession No. ML100610470.

Supporting Onsite Documentation

12. Structural Integrity Calculation 0800075.302, N-716 Evaluation for Palisades, Revision 0.

13. Structural Integrity Calculation 0800075.301 Degradation Mechanism Evaluation for Palisades, Revision 0.

Page 23 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

The degradation mechanism identified for the FPS piping is pitting, and the examination will be performed in accordance with the owner's existing localized corrosion program. The examination volume shall include base metal, welds, and weld HAZ in the affected regions of carbon and low-alloy steel, and the welds and weld HAZ of austenitic steel. Examinations shall verify the minimum wall thickness required. The examination method and examination region shall be sufficient to characterize the extent of the element degradation.

5.

REFERENCES/DOCUMENTATION

1. EPRI Report 1006937, Extension of EPRI Risk Informed lSI Methodology to Break Exclusion Region Programs.

2. EPRI TR-112657, Revised Risk-Informed Inservice Inspection Evaluation Procedure, Rev. B-A.

3. ASME Code Case N-716, Alternative Piping Classification and Examination Requirements,Section XI Division 1.

4. Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis.

5. Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping.

6. Regulatory Guide 1.200, Rev 2 An Approach For Determining The Technical Adequacy Of Probabilistic Risk Assessment Results For Risk-Informed Activities.

7. USNRC Safety Evaluation for Grand Gulf Nuclear Station Unit 1, Request for Alternative GG-ISI-002-lmplement Risk-Informed lSI based on ASME Code Case N-716, dated September 21,2007. ADAMS Accession No. ML072430005

8. USNRC Safety Evaluation for DC Cook Nuclear Plant, Units 1 and 2, Risk-Informed Safety-Based lSI program for Class 1 and 2 Piping Welds, dated September 28,2007. See ADAMS Accession No. ML072620553.

9. EPRI Report 1021467 Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs.

10. Waterford-3 Safety Evaluation - See ADAMS Accession No. ML080980120.

11. Vogtle Electric Generating Plant Safety Evaluation - See ADAMS Accession No. ML100610470.

Supporting Onsite Documentation

12. Structural Integrity Calculation 0800075.302, "N-716 Evaluation for Palisades",

Revision O.

13. Structural Integrity Calculation 0800075.301 "Degradation Mechanism Evaluation for Palisades," Revision O.

Page 23 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

System 1

Weld N-71 6 Safety Significance Determination Safety Count 2

Significance RCPB SDC PWR: FW BER CDF> 1 E-6 High Low CVCS 190 V

V 108 V

V 79 V

V V

ESS 51 V

V 458 V

38 V

V FWS 8

V V

95 V

4 V

V 4

V V

V MSS 167 V

V 51 V

229 V

V Pcs 12 V

V V

RWS 8

V V

SWS 17 V

VAS 4

V 535 V

V 91 V

V V

51 V

V Summary Results 38 V

V all 12 V

V Systems V

V V

167 V

V 625 V

(1)

Systems:

CVCS = Chemical Volume Control System ESS = Engineered Safeguards System FWS = Feedwater System MSS = Main Steam System PCS = Primary Coolant System RWS = Radwaste System SWS = Service Water System VAS = Vent & Air Conditioning System (2)

HSS non-ASME piping is not included in the Weld Count or Summary Results, except for 12 non-ASME BER welds in FWS and MSS.

Table 3.1 Code Case N-71 6 Safety Significance Determination Totals 1523 Page 24 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table 3.1 CdC N 716 S f S*"f 0

o e ase -

a ety Igm Icance etermmatlon Safety Weld N-716 Safety Significance Determination System(1)

Count(2)

Significance RCPB SOC PWR:FW BER CDF> 1E-6 High Low cves 190

-/

108

-/

79

-/

-/

ESS 51

-/

458 38 FWS 8

95 4

4 MSS 167 51 229

-/

PCS 12

-/

-/

RWS 8

-/

SWS 17 VAS 4

535

-/

91

-/

-/

Summary 51

-/

Results 38 all 12 Systems 4

167 625 Totals 1523 (1) Systems:

CVCS = Chemical Volume Control System ESS = Engineered Safeguards System FWS = Feedwater System MSS = Main Steam System PCS = Primary Coolant System RWS = Radwaste System SWS = Service Water System VAS = Vent & Air Conditioning System

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

-/

(2) HSS non-ASME piping is not included in the Weld Count or Summary Results, except for 12 non-ASME BER welds in FWS and MSS.

Page 24 of 30

-/

-/

-/

-/

-/

-/

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table 3.2 Failure Potential Assessment Summarv 2

Thermal Localized Flow Fati ue Stress_Corrosion_Cracking Corrosion Sensitive System TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC CVCS

- Chemical Volume & Control System I

ESS

- Engineered Safeguards System I

FWS Feedwater System I

MSS

- Main Steam System PCS

- Primary Coolant System F

I RWS

- Radwaste System SWS

- Service Water System VAS

- Vent & Air Conditioning System Notes:

1.

Systems are described in Table 3.1 Notes.

2.

A degradation mechanism assessment was not perlormed on low safety significant piping segments. This includes the SWS and VAS systems in their entirety, as well as portions of the ESS, FWS and MSS systems.

Page 25 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table 3.2 Thermal Localized Fati~ue Stress Corrosion Cracking Corrosion System(1)

TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CVCS - Chemical Volume & Control System ESS - Engineered Safeguards System FWS - Feedwater System MSS - Main Steam System PCS - Primary Coolant System RWS - Radwaste System SWS - Service Water System VAS - Vent & Air Conditicming System Notes:

1.

Systems are described in Table 3.1 Notes.

Flow Sensitive CC E-C FAC

2.

A degradation mechanism assessment was not performed on low safety significant piping segments. This includes the SWS and VAS systems in their entirety, as well as portions of the ESS, FWS and MSS systems.

Page 25 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table 3.3: Code Case N716 Selections Notes:

Weld 1

Count 2

N716 Selection Considerations SystemU RCPB Selections HSS LSS DMs RCPB RCPB (lFlV)

(OC)

CVCS 3

TT V

v(3) 3 cvcs 31 TT V

0 cvcs 156 None V

6 ESS 15 IGSCC V

4 ESS 4

None V

V(3)

ESS 168 None V

7 ESS 51 None 0

ESS 458 N/A 0

ws 38 None 4

ws 8

None V

1 ws 95 N/A 0

MSS 167 None 11 MSS 8

None V

8 MSS 51 N/A 0

cs 1

TASCS V

V 1

PCS 7

TASCS,TT V

V 3

pcs 3

TASCS,TT,PWSCC V

V 0

PCS 18 TT V

V 10 PCS 9

TT V

7 PCS 7

TT,PWSCC V

V o

PCS 190 None V

V 26 pcs 6

None V

0 RWS 6

None V

V 1

RWS 2

None V

0 sws 17 N/A 0

VAS 4

N/A 0

1 TASCS V

V 1

7 TASCSTT V

V 3

3 TASCS,TT,PWSCC V

V 0

21 TT V

V 13 40 TT V

7 7

TT,PWSCC V

V 0

15 IGSCC V

4 200 None V

V 31 332 None V

13 16 None V

9 256 None 4

Totals 898 625 96 (1)

Systems are described in Table 3.1 Notes.

(2)

HSS non-ASME piping is not included in the Weld Count or Summary Results except for 12 non-ASME BER welds in FWS and MSS.

(3)

Since there are only 3 IFIV welds available in the CVCS, and only 4 IFIV welds available in the ESS, PCS system welds on branch lines to the CVCS and ESS lines are used to meet the 10% HSS selection criteria.

Page 26 of 30 Notes:

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Table 3.3: Code Case N716 Selections Weld N716 Selection Considerations System(1)

Counr2)

Selections RCPB HSS LSS OMs RCPB RCPB (IFIV)

(OC)

BER eves 3

TT

./

./(3) 3 eves 31 TT

./

0 eves 156 None

./

6 ESS 15 IGSee

./

4 ESS 4

None

./

./(3) 4 ESS 168 None

./

7 ESS 51 None 0

ESS 458 N/A 0

FWS 38 None 4

FWS 8

None

./

1 FWS 95 N/A 0

MSS 167 None 11 MSS 8

None

./

8 MSS 51 N/A 0

pes 1

TASeS

./

./

1 pes 7

TASeS,TT

./

./

3 pes 3

TASeS,TT,PWSee

./

./

0 pes 18 TT

./

./

10 pes 9

TT

./

7 pes 7

TT,PWSee

./

./

0 pes 190 None

./

./

26 pes 6

None

./

0 RWS 6

None

./

./

1 RWS 2

None

./

0 SWS 17 N/A 0

VAS 4

N/A 0

1 TASeS

./

./

1 7

TASeS,TT

./

./

3 3

TASeS,TT,PWSee

./

./

0 21 TT

./

./

13 40 TT

./

7 7

TT,PWSee

./

./

0 15 IGSee

./

4 200 None

./

./

31 332 None

./

13 16 None

./

9 256 None 4

Totals 898 625 96 (1)

Systems are described in Table 3.1 Notes.

(2)

HSS non-ASME piping is not included in the Weld Count or Summary Results except for 12 non-ASME BER welds in FWS and MSS.

(3)

Since there are only 3 IFIV welds available in the eves, and only 4 IFIV welds available in the ESS, pes system welds on branch lines to the eves and ESS lines are used to meet the 10% HSS selection criteria.

Page 26 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table 3.4: Risk Impact Analysis Results Safety Break Failure Potential Inspections CDF_Impact LERF Impact System Significance Location DMs Rank SXI RIS_B Delta wIPOD wlo POD wIPOD wlo POD CVCS High LOCA TT Medium 0

3 3

-4.86E-09

-2.70E-09

-4.86E-10

-2.70E-10 CVCS High PLOCMLOCA TT Medium 0

0 0

0.OOE+00 0.OOE+00 0.OOE÷00 0.OOE÷00 CVCS High PLOCA/ILOCA None Low 0

0 0

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 CVC Total

-4.86E-09

-2.70E-09

-4.86E-1O

-2.70E-1O ESS High PLOCA/ILOCA IGSCC Medium 0

4 4

-4.OOE-11

-4.OOE-11

-4.OOE-12

-4.OOE-12 ESS High LOCA None Low 0

0 0

0.OOE+00 0.OOE+00 OOOE+OO O.OOE÷00 ESS High PLOCAJILOCA None Low 0

0 0

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 ESS High PPLOCA None Low 0

0 0

0.OOE+00 0.OOE+00 O.OOE+OO 0.OOE+00 ESS Low Class 2 LSS 11 0

-11 1.1OE-1O 1.1OE-1O 1.1OE-11 1.1OE-11 ESS Total 7.OOE-11 7.OOE-11 7.OOE-12 7.OOE-12 FWS High MSLB-l None Low 4

4 0

0.OOE+00 0.OOE÷00 0.OOE+00 0.OOE÷00 FWS High MSLB-0 None Low 0

1 1

-5.OOE-11

-5.OOE-11

-5.OOE-12

-5.OOE-12 FWS Low MSLB-l 16 0

-16 1.60E-1O 1.60E-10 1.60E-11 1.60E-11 FWS Low MSLB-0 2

0

-2 2.OOE-09 2.OOE-09 2.OOE-10 2.OOE-1O FWS Total 1.1OE-1O 1.1OE-10 1.1OE-11 1.1OE-11 MSS High MSLB-0 None Low 10 19 9

-4.50E-10

-4.50E-1O

-4.50E-11

-4.50E-11 MSS Low MSLB-l 0

0 0

0.OOE+00 O.OOE-i-OO O.OOE-t-O0 0.OOE+00 MSS Total

-4.50E-1O

-4.50E-1O

-4.50E-11

-4.50E-11 PCS High LOCA TASCS Medium 0

1 1

-1.62E-09

-9.OOE-1O

-1.62E-10

-9.OOE-11 PCS High LOCA TASCSTT Medium 0

3 3

-4.86E-09

-270E-09

-4.86E-1O

-2.70E-10 PCS High LOCA TASCS,TT, Medium 3

0

-3 2.70E-09 2.70E-09 2.70E-10 2.70E-10 PCS High LOCA TT Medium 0

10 10

-1.62E-08

-9.OOE-09

-1.62E-09

-9.OOE-10 PCS High PLOCMLOCA TT Medium 0

7 7

-1.26E-10

-7.OOE-11

-1.26E-11

-7.OOE-12 PCS High LOCA TT,PWSCC Medium 2

0

-2 1.80E-09 1.80E-09 1.80E-10 1.80E-10 PCS High LOCA None Low 30 21

-9 4.05E-10 4.05E-1O 4.05E-11 4.05E-11 PCS High PLOCA/ILOCA None Low 0

0 0

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 PCS Total

-1.79E-08

-7.77E-09

-1.79E-09

-7.77E-1O Page 27 of 30 Safety System Significance CVCS High CVCS High CVCS High CVCTotal ESS High ESS High ESS High ESS High ESS Low ESS Total FWS High FWS High FWS Low FWS Low FWS Total MSS High MSS Low MSS Total PCS High PCS High PCS High PCS High PCS High PCS High PCS High PCS High PCS Total ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Table 3.4: Risk Impact Analysis Results Break Failure Potential Inspections CDF Impact Location OMs Rank SXI RIS_B Delta wlPOD w/oPOD LOCA IT Medium 0

3 3

-4.B6E-09

-2.70E-09 PLOCAlILOCA IT Medium 0

0 0

O.OOE+OO O.OOE+OO PLOCAlILOCA None Low 0

0 0

O.OOE+OO O.OOE+OO

-4.S6E-09

-2.70E-09 PLOCAlILOCA IGSCC Medium 0

4 4

-4.00E-11

-4.00E-11 LOCA None Low 0

0 0

O.OOE+OO O.OOE+OO PLOCAlILOCA None Low 0

0 0

O.OOE+OO O.OOE+OO PPLOCA None Low 0

0 0

O.OOE+OO O.OOE+OO Class 2 LSS Assume 11 0

-11 1.10E-10 1.10E-10 Medium 7.00E-11 7.00E-11 MSLB-I None Low 4

4 0

O.OOE+OO O.OOE+OO MSLB-O None Low 0

1 1

-S.00E-11

-S.00E-11 MSLB-I Assume 16 0

-16 1.60E-10 1.60E-10 Medium MSLB-O Assume 2

0

-2 2.00E-09 2.00E-09 Medium 1.10E-10 1.10E-10 MSLB-O None Low 10 19 9

-4.S0E-10

-4.S0E-10 MSLB-I Assume 0

0 0

O.OOE+OO O.OOE+OO Medium

-4.50E-10

-4.50E-10 LOCA TASes Medium 0

1 1

-1.62E-09

-9.00E-10 LOCA TASCS,IT Medium 0

3 3

-4.B6E-09

-2.70E-09 LOCA TASCS,IT, Medium 3

0

-3 2.70E-09 2.70E-09 PWSCC LOCA IT Medium 0

10 10

-1.62E-OB

-9.00E-09 PLOCAlILOCA IT Medium 0

7 7

-1.26E-10

-7.00E-11 LOCA IT,PWSCC Medium 2

0

-2 1.BOE-09 1.BOE-09 LOCA None Low 30 21

-9 4.0SE-10 4.0SE-10 PLOCAIILOCA None Low 0

0 0

O.OOE+OO O.OOE+OO

-1.79E-oS

-7.77E-Q9 Page 27 of 30 LERF Impact wlPOD wlo POD

-4.B6E-10

-2.70E-10 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO

-4.S6E-10

-2.70E-10

-4.00E-12

-4.00E-12 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO 1.10E-11 1.10E-11 7.00E-12 7.00E-12 O.OOE+OO O.OOE+OO

-S.00E-12

-S.OOE-12 1.60E-11 1.60E-11 2.00E-10 2.00E-10 1.10E-11 1.10E-11

-4.S0E-11

-4.S0E-11 O.OOE+OO O.OOE+OO

-4.50E-11

-4.50E-11

-1.62E-10

-9.00E-11

-4.B6E-10

-2.70E-10 2.70E-10 2.70E-10

-1.62E-09

-9.00E-10

-1.26E-11

-7.00E-12 1.BOE-10 1.BOE-10 4.0SE-11 4.OSE-11 O.OOE+OO O.OOE+OO

-1.79E-Q9

-7.77E-10

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Safety Break Failure Potential Inspections CDF_Impact LERF Impact System Significance Location DMs Rank SXI RIS_B Delta wIPOD wlo POD w/POD wlo POD RWS High LOCA None Low 0

0 0

O.OOE+00 0.OOE+00 O.OOE+OO O.OOE+OO RWS High PLOCMLOCA None Low 0

0 0

O.OOE+O0 O.OOE+OO O.OOE+00 O.OOE+OO RWS Total O.OOE+OO O.OOE÷OO O.OOE+OO O.OOE+OO Assume SWS Total Low Class 2 LSS 3

0

-3 3.OOE-11 3.OOE-11 3.OOE-12 3.OOE-12 Medium Assume VAS Total Low Class 2 LSS 0

0 0

O.OOE+OO O.OOE+OO O.OOE÷OO O.OOE÷OO Medium Grand Total 81 73

-8 2.30E08 1.07E08 2.30E09 1.07E09 Notes 1.

Systems are described in Table 3.1 Notes.

2.

Only those ASME Section Xl Code inspection locations that received a volumetric examination are included in the count. Inspection locations previously subjected to a surface examination only were not considered in accordance with Section 3.7.1 of EPRI TR-112657.

3.

Only those RIS_B inspection locations that receive a volumetric examination are included in the count. Locations subjected to VT2 only are not credited in the count for risk impact assessment.

4.

The failure potential rank for high safety significant (HSS) locations is assigned as High, Medium, or Low depending upon potential susceptibly to the various types of degradation. [Note: Low Safety Significant (LSS) locations were conservatively assumed to be a rank of Medium (i.e., Assume Medium) 5.

The LSS designation is used to identify those Code Class 2 locations that are not HSS because they do not meet any of the five HSS criteria of Section 2(a) of N-716 (e.g., not part of the BER scope).

6.

As described in Section 2.2, Alloy 82/182 welds susceptible to no degradation mechanism or PWSCC only per the RIS_B Program failure potential assessment are not included in the table.

7.

The risk reduction associated with the non-ASME piping is not included in the above table, except for 12 non-ASME BER welds in FWS and MSS.

Page 28 of 30 Safety System Significance RWS High RWS High RWS Total SWSTotal Low VAS Total Low Grand Total Notes ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a{a){3){i)

Break Failure Potential Inspections CDF Impact Location DMs Rank SXI RIS_B Delta wlPOD wlo POD LOCA None Low 0

0 0

O.OOE+OO O.OOE+OO PLOCAlILOCA None Low 0

0 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO Class 2 LSS Assume 3

0

-3 3.00E-11 3.00E-11 Medium Class 2 LSS Assume 0

0 0

O.OOE+OO O.OOE+OO Medium 81 73

-8

-2.30E-08

-1.07E-Q8

1.

Systems are described in Table 3.1 Notes.

LERFlmpact wlPOD wlo POD O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO 3.00E-12 3.00E-12 O.OOE+OO O.OOE+OO

-2.30E-09

-1.07E-09

2.

Only those ASME Section XI Code inspection locations that received a volumetric examination are included in the count. Inspection locations previously subjected to a surface examination only were not considered in accordance with Section 3.7.1 of EPRI TR-112657.

3.

Only those RIS_B inspection locations that receive a volumetric examination are included in the count. Locations subjected to VT2 only are not credited in the count for risk impact assessment.

I

4.

The failure potential rank for high safety significant (HSS) locations is assigned as "High", "Medium", or "Low" depending upon potential susceptibly to the various types of degradation. [Note: Low Safety Significant (LSS) locations were conservatively assumed to be a rank of Medium (i.e., "Assume Medium")

5.

The "LSS" designation is used to identify those Code Class 2 locations that are not HSS because they do not meet any of the five HSS criteria of Section 2(a) of N-716 (e.g., not part of the BER scope).

6.

As described in Section 2.2, Alloy 82/182 welds susceptible to no degradation mechanism or PWSCC only per the RIS_B Program failure potential assessment are not included in the table.

7.

The risk reduction associated with the non-ASME piping is not included in the above table, except for 12 non-ASME BER welds in FWS and MSS.

Page 28 of 30

Notes 1.

Systems are described in Table 3.1 Notes.

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table 4: Inspection Location Selections Comparison S stem Significance Break Failure Potential Code Weld Section Xl Code Case N716

Location Category Count High Low DMs Rank Vol Surface RIS_B Other CVCS V

LOCA TT Medium B-J 3

0 0

3 0

CVCS V

PLOCA/ILOCA TT Medium B-J 31 0

3 0

0 CVCS V

PLOCA/ILOCA None Low B-J 156 0

11 0

6 ESS V

PLOCA/ILOCA IGSCC Medium B-J 15 0

0 4

0 ESS V

LOCA None Low B-J 4

0 1

0 4

ESS V

PLOCA/ILOCA None Low B-J 168 0

10 0

7 ESS V

PPLOCA None Low C-F-i 51 0

0 0

0 ESS V

Class 2 LSS Assume Medium C-F-i 458 11 ii 0

0 FWS V

MSLB-l None Low C-F-2 28 4

0 4

0 FWS V

MSLB-O None Low C-F-2, Aug 18 0

0 1

0 FWS V

MSLB-l Assume Medium C-F-2 76 16 0

0 0

FWS V

MSLB-O Assume Medium C-F-2 19 2

0 0

0 MSS V

MSLB-O None Low C-F-2, Aug 175 10 0

19 0

MSS V

MSLB-I Assume Medium C-F-2 51 0

1 0

0 PCS V

LOCA TASCS Medium B-J 1

0 0

1 0

PCS V

LOCA TASCS,TT Medium B-J 7

0 0

3 0

PCS V

LOCA TASCS,TT, PWSCC Medium B-J 3

3 0

0 0

PCS V

LOCA TT Medium B-J 18 0

10 10 0

PCS V

PLOCMLOCA TT Medium B-J 9

0 0

7 0

PCS V

LOCA TT,PWSCC Medium B-F, B-J 7

2 1

0 0

PCS V

LOCA None Low B-J 190 30 10 21 5

PCS V

PLOCA/ILOCA None Low B-J 6

0 1

0 0

RWS V

LOCA None Low B-J 6

0 0

0 1

RWS V

PLOCAIILOCA None Low B-J 2

0 0

0 0

SWS V

Class 2 LSS Assume Medium C-F-2 17 3

0 0

0 VAS V

Class 2 LSS Assume Medium C-F-2 4

0 0

0 0

Totals 1523 81 59 73 23 Page 29 of 30 Safety System Significance High Low CVCS 0/

CVCS 0/

CVCS 0/

ESS 0/

ESS 0/

ESS 0/

ESS 0/

ESS 0/

FWS 0/

FWS 0/

FWS 0/

FWS 0/

MSS 0/

MSS 0/

PCS 0/

PCS 0/

PCS 0/

PCS 0/

PCS 0/

PCS 0/

PCS 0/

PCS 0/

RWS 0/

RWS 0/

SWS 0/

VAS 0/

Notes ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.SSa(a)(3)(i)

Tabl L

Sel C

Break Failure Potential Code Weld Section XI Location OMs Rank Category Count Vol Surface lOCA TT Medium B-J 3

0 0

PlOCAlllOCA TT Medium B-J 31 0

3 PlOCAlllOCA None low B-J 156 0

11 PlOCAlllOCA IGSCC Medium B-J 15 0

0 lOCA None low B-J 4

0 1

PlOCAlllOCA None low B-J 168 0

10 PPlOCA None low C-F-1 51 0

0 Class2lSS Assume Medium C-F-1 458 11 11 MSlB-1 None low C-F-2 28 4

0 MSlB-O None low C-F-2, Aug 18 0

0 MSlB-1 Assume Medium C-F-2 76 16 0

MSlB-O Assume Medium C-F-2 19 2

0 MSlB-O None low C-F-2, Aug 175 10 0

MSlB-1 Assume Medium C-F-2 51 0

1 lOCA TASCS Medium B-J 1

0 0

lOCA TASCS,TT Medium B-J 7

0 0

lOCA TASCS,TT, PWSCC Medium B-J 3

3 0

lOCA TT Medium B-J 18 0

10 PlOCAlllOCA TT Medium B-J 9

0 0

lOCA TT,PWSCC Medium B-F, B-J 7

2 1

lOCA None low B-J 190 30 10 PlOCAlllOCA None low B-J 6

0 1

lOCA None low B-J 6

0 0

PlOCAlllOCA None low B-J 2

0 0

Class 2lSS Assume Medium C-F-2 17 3

0 Class2lSS Assume Medium C-F-2 4

0 0

Totals 1523 81 59

1.

Systems are described in Table 3.1 Notes.

Page 29 of 30 Code Case N716 RIS_B Other 3

0 0

0 0

6 4

0 0

4 0

7 0

0 0

0 4

0 1

0 0

0 0

0 19 0

0 0

1 0

3 0

0 0

10 0

7 0

0 0

21 5

0 0

0 1

0 0

0 0

0 0

73 23

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 2.

The column labeled Other is generally used to identify plant augmented inspection program locations credited per Section 4 of Code Case N-716. Code Case N-716 allows the existing plant augmented inspection program for IGSCC (Categories B through G) in a BWR to be credited toward the 10%

requirement. This option is not applicable for the Palisades RIS_B application. The Other column has been retained in this table solely for uniformity purposes with other RIS_B application template submittals and to indicate when RIS_B selections will receive a VT-2 examination (these are not credited in risk impact assessment).

3.

The failure potential rank for high safety significant (HSS) locations is assigned as High, Medium, or Low depending upon potential susceptibly to the various types of degradation. [Note: Low safety significant (LSS) locations were conservatively assumed to be a rank of Medium (i.e., Assume Medium).

4.

As described in Section 2.2, Alloy 82/182 welds susceptible to no degradation mechanism or PWSCC only per the RIS_B Program failure potential assessment are not included in the table.

5.

inspection locations associated with the HSS non-ASME piping are not included in the above table except for 12 non-ASME BER welds in FWS and MSS.

Page 30 of 30 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

2.

The column labeled "Other" is generally used to identify plant augmented inspection program locations credited per Section 4 of Code Case N-716. Code Case N-716 allows the existing plant augmented inspection program for IGSCC (Categories B through G) in a BWR to be credited toward the 10%

requirement. This option is not applicable for the Palisades RIS_B application. The "Other" column has been retained in this table solely for uniformity purposes with other RIS_B application template submittals and to indicate when RIS_B selections will receive a VT-2 examination (these are not credited in risk impact assessment).

3.

The failure potential rank for high safety significant (HSS) locations is assigned as "High", "Medium", or "low" depending upon potential susceptibly to the various types of degradation. [Note: low safety significant (LSS) locations were conservatively assumed to be a rank of Medium (Le., "Assume Medium").

4.

As described in Section 2.2, Alloy 82/182 welds susceptible to no degradation mechanism or PWSCC only per the RIS_B Program failure potential assessment are not included in the table.

5.

Inspection locations associated with the HSS non-ASME piping are not included in the above table except for 12 non-ASME BER welds in FWS and MSS.

Page 30 of 30

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Attachment A to Palisades N-716 Template Consideration of the Adequacy of Probabilistic Risk Assessment Model for Application of Code Case N716 Palisades PRA Response to RG 1.200 Peer Review Findings 55 Pages Follow ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a}(3}(i}

Attachment A to Palisades N-716 Template Consideration of the Adequacy of Probabilistic Risk Assessment Model for Application of Code Case N716 Palisades PRA Response to RG 1.200 Peer Review Findings 55 Pages Follow

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

PRA Quality This attachment describes the PRA model peer review history and resolution of peer review findings and observations (F&Os) between completion of the previous PRA model analysis of record, PSAR2c [24] and the PRA flooding analysis in December of 2013 [43].

In addition, historical information is provided regarding model reviews performed prior to the issuance of RG 1.200.

A2.1 Objective and Scope 2

A2.2 Conclusion 2

A2.3 Combustion Engineering Owners Group (CEOG) Peer Review (2000) 3 A2.4 Gap Analysis (2004) 3 A2.5 Full Power Internal Events Peer Review 5

A2.6 Refrences 54 Page 1 of 55 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

PRA Quality This attachment describes the PRA model peer review history and resolution of peer review findings and observations (F&Os) between completion of the previous PRA model analysis of record, PSAR2c [24] and the PRA flooding analysis in December of 2013 [43]. In addition, historical information is provided regarding model reviews performed prior to the issuance of RG 1.200.

A2.1 Objective and Scope.........*.*.*.................*.**............................................*........................... 2 A2.2 Conclusion........*.......*......*.*...*...................**........................................................................ 2 A2.3 Combustion Engineering Owners Group (CEOG) Peer Review (2000)............................. 3 A2.4 Gap Analysis (2004).........***.*.............................................................................................. 3 A2.S Full Power Internal Events Peer Review............................................................................. 5 A2.6 Refrences.................*.......*..............................................*................................................. 54 Page 1 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

A2.1 OBJECTIVE AND SCOPE This purpose of this attachment is to describe the PRA model peer review history and resolution of peer review findings and observations (F&Os) between completion of the previous PRA model analysis of record, PSAR2c [24] and development of the flooding analysis completed in December 2013 [36].

In addition to the internal events peer review results described here, portions of Palisades Nuclear Plant (PNP) model used for this update underwent a 3-phase fire PRA peer review during its development.

Phase 1 Initial in-process peer review performed January 18, 2010 Phase 2 Second in-process peer review performed August 22, 2010 Final Final peer review performed March 20, 2011 The fire PRA peer review report is provided in SCIENTECH document 17825-1 [33]. Resolutions to the fire PRA F&Os may be found in Attachment 0 of Reference [16].

A

2.2 CONCLUSION

The resolution of each of the peer review team findings and observations documented in Tables A2.4-1 and A2.5-1, demonstrate that the Palisades Full Power Internal Events PRA model meets at least capability category II in all Regulatory Guide 1.200 standard supporting requirements with some exceptions.

Four of the fifty-two FPIE findings have not been addressed. Findings QU-Ci-Ol and HR-G7-01 related to human error dependency analysis have not been completed. Finding QU-D1 -01, related to the final formal documentation and model issuance is incomplete. Finding QU-B2-01 related to establishing the final truncation limit for the full power internal events model is also incomplete. Closure of these last four findings is dependent on completion of the final human error dependency analysis followed by final documentation of the model results and the truncation study. The remaining 48 findings and 26 suggestions have been resolved. Flooding PRA specific human error dependency and truncation studies were completed for this application.

The only findings that remain open are associated with the full power internal events PRA.

Page 2 of 55 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a}(3}(i}

A2.1 OBJECTIVE AND SCOPE This purpose of this attachment is to describe the PRA model peer review history and resolution of peer review findings and observations (F&Os ) between completion of the previous PRA model analysis of record, PSAR2c [24] and development of the flooding analysis completed in December 2013 [36].

In addition to the internal events peer review results described here, portions of Palisades Nuclear Plant (PNP) model used for this update underwent a 3-phase fire PRA peer review during its development.

Phase 1 -

Initial in-process peer review performed January 18, 2010 Phase 2 - Second in-process peer review performed August 22, 2010 Final - Final peer review performed March 20, 2011 The fire PRA peer review report is provided in SCIENTECH document 17825-1 [33]. Resolutions to the fire PRA F&Os may be found in Attachment 0 of Reference [16].

A

2.2 CONCLUSION

The resolution of each of the peer review team findings and observations documented in Tables A2.4-1 and A2.5-1, demonstrate that the Palisades Full Power Internal Events PRA model meets at least capability category" in all Regulatory Guide 1.200 standard supporting requirements with some exceptions.

Four of the fifty-two FPIE findings have not been addressed. Findings QU-C1-01 and HR-G7-01 related to human error dependency analysis have not been completed. Finding QU-D1-01,

related to the final formal documentation and model issuance is incomplete. Finding QU-B2-01 related to establishing the final truncation limit for the full power internal events model is also incomplete. Closure of these last four findings is dependent on completion of the final human error dependency analysis followed by final documentation of the model results and the truncation study. The remaining 48 findings and 26 suggestions have been resolved. Flooding PRA specific human error dependency and truncation studies were completed for this application.

The only findings that remain open are associated with the full power internal events PRA.

Page 2 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

A2.3 COMBUSTION ENGINEERING OWNERS GROUP (CEOG) PEER REVIEW (2000)

The CEOG conducted an industry peer review of the Palisades PRA in 2000 [2]. All level A and B findings have been addressed.

A2.4 GAP ANALYSIS (2004)

Subsequent to the 2000 peer review, a gap analysis was performed in 2004 [3].

At the behest of the NRC, the industry undertook a task to develop a consensus standard on the technical adequacy of PRAs for regulatory applications. This effort resulted in publication of ASME RA-S-2002. Concurrently, under the direction of the Nuclear Energy Institute (NEI) and the Owners Groups for each major reactor provider, peer reviews of PRA5 were conducted using the guidance in NEI 00-02. The NRC was also concurrently developing guidance for determining the adequacy of risk analyses for use in regulatory applications. The first draft of this guidance was published as Draft Guide 1122 (DG-1122) in September 2002. Following interactions with industry in subsequent years as the ASME Standard was being modified, the NRC published DG-1 161 in September 2006. This draft version of Regulatory Guide 1.200 (RG 1.200) provided guidance on self-assessments to determine the adequacy of PRAs.

This assessment reviewed the peer review facts against the guidance in DG-i 122 and produced a list of recommended actions to address gaps between the results of the peer review and the guidance in DG-1122. As noted above, Palisades had subsequently addressed all A and B level facts and observations (F&Os) from the peer review certification report. DG-i 122 allowed for two mechanisms for conducting a self-assessment. One was a direct comparison of the PRA against the Standard with additional considerations cited by the NRC to address areas where the NRC did not agree with the Standard (Table A-i of DG-1 122). The other method was to take advantage of the peer review findings and perform additional reviews against the Standard in areas where the NRC found that the peer review process needed additional effort to address NRC concerns with the Standard. The NRC issues were documented in Table B-4 of DG-1122.

This was the method used in the Palisades Gap Analysis.

Table A2.4-i lists the recommended actions identified by this evaluation. In general, the additional recommendations deal with issues of documentation and/or justification for technical analyses in the PRA. Slightly less than half of the additional recommendations potentially resulted in a change to the actual model. Only three additional recommendations potentially resulted in a noticeable change in the CDF or LERF. These included the removal of EDG repair from the model, the inclusion of additional flow diversion paths for key systems, and the inclusion of potential concurrent unavailabilities (such as train wise maintenance schedules where one train in multiple systems is taken out of service at the same time. These issues have all been subsequently addressed in earlier model updates.

Table A2.4-i Gap Analysis Recommendations tern Description of Issues Applicable SR Model Changes Disposition Numbers Needed?

Document the rationale for not using precursors to IE-A7 No Addressed A

identify initiators.

Walkdowns/interviews with operators and engineers B

have been conducted in the past, but need to be done SY-A4, IF-B3, No Addressed again in light of recent PRA updates and staffing IF-C8, IF-E8 changes.

Page 3 of 55 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

A2.3 COMBUSTION ENGINEERING OWNERS GROUP (CEOG) PEER REVIEW (2000)

The CEOG conducted an industry peer review of the Palisades PRA in 2000 [2]. All level A and 8 findings have been addressed.

A2.4 GAP ANALYSIS (2004)

Subsequent to the 2000 peer review, a gap analysis was performed in 2004 [3].

At the behest of the NRC, the industry undertook a task to develop a consensus standard on the technical adequacy of PRAs for regulatory applications. This effort resulted in publication of ASME RA-S-2002. Concurrently, under the direction of the Nuclear Energy Institute (NEI) and the Owners Groups for each major reactor provider, peer reviews of PRAs were conducted using the guidance in NEI 00-02. The NRC was also concurrently developing guidance for determining the adequacy of risk analyses for use in regulatory applications. The first draft of this guidance was published as Draft Guide 1122 (DG-1122) in September 2002. Following interactions with industry in subsequent years as the ASME Standard was being modified, the NRC published DG-1161 in September 2006. This draft version of Regulatory Guide 1.200 (RG 1.200) provided guidance on self-assessments to determine the adequacy of PRAs.

This assessment reviewed the peer review facts against the guidance in DG-1122 and produced a list of recommended actions to address "gaps" between the results of the peer review and the guidance in DG-1122. As noted above, Palisades had subsequently addressed all A and 8 level facts and observations (F&Os) from the peer review certification report. DG-1122 allowed for two mechanisms for conducting a self-assessment. One was a direct comparison of the PRA against the Standard with additional considerations cited by the NRC to address areas where the NRC did not agree with the Standard (Table A-1 of DG-1122). The other method was to take advantage of the peer review findings and perform additional reviews against the Standard in areas where the NRC found that the peer review process needed additional effort to address NRC concerns with the Standard. The NRC issues were documented in Table 8-4 of DG-1122.

This was the method used in the Palisades Gap Analysis.

Table A2.4-1 lists the recommended actions identified by this evaluation. In general, the additional recommendations deal with issues of documentation and/or justification for technical analyses in the PRA. Slightly less than half of the additional recommendations potentially resulted in a change to the actual model. Only three additional recommendations potentially resulted in a noticeable change in the CDF or LERF. These included the removal of EDG repair from the model, the inclusion of additional flow diversion paths for key systems, and the inclusion of potential concurrent unavailabilities (such as train wise maintenance schedules where one train in multiple systems is taken out of service at the same time. These issues have all been subsequently addressed in earlier model updates.

Table A2.4-1 Gap Analysis Recommendations Item Description of Issues Applicable SR Model Changes Disposition Numbers Needed?

A Document the rationale for not using "precursors" to IE-A7 No Addressed identify initiators.

Walkdownslinterviews with operators and engineers B

have been conducted in the past, but need to be done SY -A4, I F-B3, No Addressed again in light of recent PRA updates and staffing IF-C8, IF-E8 changes.

Page 3 of 55

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Table A2.4-1 Gap Analysis Recommendations Item Description of Issues Applicable SR Model Changes Numbers Needed?

Disposition Flow diversions are included in many systems but SY-A12b Yes Addressed C

additional cases need to be included in the model.

D Concurrent unavailabilities should be included in the DA-C13 Yes Addressed model.

Palisades included inter-area propagation but needs to E

include unavailability of flood barriers such as IF-C3b Yes Addressed doors/hatches.

Palisades credited flood isolation operator actions after 30 minutes. Further activity is underway to document IF-C7 No Addressed F

the time available and the reliability of the potential actions.

Generic and plant specific experience was used in determining pipe failure frequencies, but factors such as lF-D5a Yes Addressed G

the impact of FAC, water hammer, etc. should be included in the analysis.

Key assumptions were documented but key H

uncertainties in the analysis need to be documented IF-F3 No Addressed and evaluated.

ISLOCA evaluation included pressure capability of secondary systems. Capability for valve closure under IE-Cl 1 No Addressed I

high flow/dP to isolate ISLOCA was not credited.

Document the rationale for this exclusion.

The pre-initiators were identified primarily based on test and maintenance activities. Inspection activities also HR-Al No Addressed

should be addressed explicitly for potential pre initiators.

The quality of procedures and processes were examined to the extent that the THERP methodology K

calls for, but do not include all the factors in the latest HR-D3 No Addressed version of DG-l 161. Document how the pre-initiator HEPs account for the quality factors noted in DG-ll 61.

L EDG repair is the only case where repair is credited.

DA-C14 Yes Addressed Palisades intends to remove that feature from the PRA.

The flooding analysis did not consider ranges of flow rates for flood sources, but used maximum flow rates M

instead. Determine if lesser flow rates would impact the lF-B3 Yes Addressed results and include as warranted.

Barrier availability was generally not accounted for but N

reverse flow via failed check valves was included in the IF-C3b Yes Addressed flooding analysis. Include potential barrier unavailability.

CCF groups were not reduced to account for the effects of flooding. This results in pessimistic (conservative) 0 impact of CDF for flooding sequences. Document the IF-E6a No Addressed rationale for not adjusting CCF group sizes for equipment that would be failed by flooding scenarios.

Sensitivity analyses on key assumptions have been performed over time but have not been documented in a comprehensive manner. Consider referencing QU-E4 No Addressed sensitivity analyses in EA calculations in the documentation of the current version of the model and subsequent updates.

Page 4 of 55 Item C

D E

F G

H I

J K

L M

N 0

P ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.4-1 Gap Analysis Recommendations Description of Issues Applicable SR Model Changes Numbers Needed?

Flow diversions are included in many systems but SY-A12b Yes additional cases need to be included in the model.

Concurrent unavailabilities should be included in the DA-C13 Yes model.

Palisades included inter-area propagation but needs to include unavailability of flood barriers such as IF-C3b Yes doors/hatches.

Palisades credited flood isolation operator actions after 30 minutes. Further activity is underway to document IF-C7 No the time available and the reliability of the potential actions.

Generic and plant specific experience was used in determining pipe failure frequencies, but factors such as IF-DSa Yes the impact of FAC, water hammer, etc. should be included in the analysis.

Key assumptions were documented but key uncertainties in the analysis need to be documented IF-F3 No and evaluated.

ISLOCA evaluation included pressure capability of secondary systems. Capability for valve closure under IE-C11 No high flow/dP to isolate ISLOCA was not credited.

Document the rationale for this exclusion.

The pre-initiators were identified primarily based on test and maintenance activities. Inspection activities also HR-A1 No should be addressed explicitly for potential pre-initiators.

The quality of procedures and processes were examined to the extent that the THERP methodology calls for, but do not include all the factors in the latest HR-D3 No version of DG-1161. Document how the pre-initiator HEPs account for the quality factors noted in DG-1161.

EDG repair is the only case where repair is credited.

DA-C14 Yes Palisades intends to remove that feature from the PRA.

The flooding analysis did not consider ranges of flow rates for flood sources, but used maximum flow rates IF-B3 Yes instead. Determine if lesser flow rates would impact the results and include as warranted.

Barrier availability was generally not accounted for but reverse flow via failed check valves was included in the IF-C3b Yes flooding analysis. Include potential barrier unavailability.

CCF groups were not reduced to account for the effects of flooding. This results in pessimistic (conservative) impact of CDF for flooding sequences. Document the IF-E6a No rationale for not adjusting CCF group sizes for equipment that would be failed by flooding scenarios.

Sensitivity analyses on key assumptions have been performed over time but have not been documented in a comprehensive manner. Consider referencing QU-E4 No sensitivity analyses in EA calculations in the documentation of the current version of the model and subsequent updates.

Page 4 of 55 Disposition Addressed Addressed Addressed Addressed Addressed Addressed Addressed Addressed Addressed Addressed Addressed Addressed Addressed Addressed

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

A2.5 FULL POWER INTERNAL EVENTS PEER REVIEW The final report documenting the results of the full-scope Regulatory Guide (RG) 1.200 [40] peer review for Palisades Probabilistic Risk Assessment (PRA) was received on March 12, 2010.

Per Reference [4], the Palisades Nuclear Power Plant Probabilistic Risk Assessment (PRA) FPIE analyses were reviewed against the requirements of Section 2 of the ASME/American Nuclear Society (ANS) Combined PRA standard [31], and the requirements of Regulatory Guide (RG) 1.200, Revision 2 [1]. This peer review was performed using the process defined in Nuclear Energy Institute (NEI) 05-04 [32].

The report summary states:

The ASME PRA Standard (ASME/ANS RA-S-2008a) contains a total 326 numbered supporting requirements in fourteen technical elements and the configuration control element. Of the 326 SRs, thirteen were determined to be not applicable to the Palisades PRA. As shown on Table 4-1, of the 313 remaining SRs, 263 SRs, or 84%, were rated as Capability Category II or greater and about 5% were Capability Category I. Only 11% of the SRs were rated as not met.

Tables 4-2 through 4-11 present the summary of the results of this peer review for each of the fifty-six HLRs from the ASME PRA Standard. In the course of this review, eighty new Facts and Observations (F&Os) were prepared, including two Best Practices (SY-A 13-02 andASA9-01), twenty-six Suggestions and fifty-two Findings. All of the new F&Os are presented in Table 4-12.

The report concludes:

Overall, the Palisades PRA was found to substantially meet the ASME PRA Standard at Capability Category!! and can be used to support risk-hformed applications. Dependent upon the specifics of the application, additional supporting analyses may be needed, particularly for applications that impact elements with unresolved findings or where an assumption could impact the conclusions of the application.

Fourteen of the fifty-two findings and three of the twenty-six suggestions were related to internal flooding issues. The resolution of all full power internal events findings and suggestions, including those related to flooding are provided Table A2.5-1. F&Os related to flooding are presented at the beginning of the Table. The remaining FPIE F&Os are presented in alphabetical order by ASME supporting requirement.

Four of the fifty-two FPIE findings have not been addressed. Findings QU-Ci -01 and HR-G7-01 related to human error dependency analysis have not been completed. Finding QU-Di -01, related to the final formal documentation and model issuance is incomplete. Finding QU-B2-01 related to establishing the final truncation limit for the full power internal events model is also incomplete. Closure of these last four findings is dependent on completion of the final human error dependency analysis followed by final documentation of the model results and the truncation study. The remaining 48 findings and 26 suggestions have been resolved. Flooding PRA specific human error dependency and truncation studies were completed for this application.

The only findings that remain open are associated with the full power internal events PRA.

Page 5 of 55 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

A2.5 FULL POWER INTERNAL EVENTS PEER REVIEW The final report documenting the results of the full-scope Regulatory Guide (RG) 1.200 [40] peer review for Palisades Probabilistic Risk Assessment (PRA) was received on March 12, 2010.

Per Reference [4], the Palisades Nuclear Power Plant Probabilistic Risk Assessment (PRA) FPIE analyses were reviewed against the requirements of Section 2 of the ASME/American Nuclear Society (ANS) Combined PRA standard [31], and the requirements of Regulatory Guide (RG) 1.200, Revision 2 [1]. This peer review was performed using the process defined in Nuclear Energy Institute (NEI) OS-04 [32].

The report summary states:

liThe ASME PRA Standard (ASMEIANS RA-S-2008a) contains a total 326 numbered supporting requirements in fourteen technical elements and the configuration control element. Of the 326 SRs, thirteen were determined to be not applicable to the Palisades PRA. As shown on Table 4-1, of the 313 remaining SRs, 263 SRs, or 84%, were rated as Capability Category II or greater and about 5% were Capability Category I. Only 11% of the SRs were rated as not met.

Tables 4-2 through 4-11 present the summary of the results of this peer review for each of the fifty-six HLRs from the ASME PRA Standard. In the course of this review, eighty new Facts and Observations (F&Os) were prepared, including two "Best Practices" (SY-A 13-02 and ASA9-01), twenty-six "Suggestions" and fifty-two IIFindings". All of the new F&Os are presented in Table 4-12."

The report concludes:

IIOverall, the Palisades PRA was found to substantially meet the ASME PRA Standard at Capability Category /I and can be used to support risk-informed applications. Dependent upon the specifics of the application, additional supporting analyses may be needed, particularly for applications that impact elements with unresolved findings or where an assumption could impact the conclusions of the application."

Fourteen of the fifty-two findings and three of the twenty-six suggestions were related to internal flooding issues. The resolution of all full power internal events findings and suggestions, including those related to flooding are provided Table A2.S-1. F&Os related to flooding are presented at the beginning of the Table. The remaining FPIE F&Os are presented in alphabetical order by ASME supporting requirement.

Four of the fifty-two FPIE findings have not been addressed. Findings QU-C1-01 and HR-G7-01 related to human error dependency analysis have not been completed. Finding QU-D1-01, related to the final formal documentation and model issuance is incomplete. Finding QU-B2-01 related to establishing the final truncation limit for the full power internal events model is also incomplete. Closure of these last four findings is dependent on completion of the final human error dependency analysis followed by final documentation of the model results and the truncation study. The remaining 48 findings and 26 suggestions have been resolved. Flooding PRA specific human error dependency and truncation studies were completed for this application.

The only findings that remain open are associated with the full power internal events PRA.

Page 5 of 55

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Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition (Supporting Suggestion Category II Text Requirement) lEEV-A5-01 Finding DETERMINE the flood-Key-words are used to identify the potential applicable This finding is related to the development of the overall plant flood initiating event frequency for LERs in the INPO Database (5 found). However, as frequency and its subsequent use to determine the maintenance each flood scenario group noted by the additional LERs identified in 5750 (3 contribution to flood frequency.

by using the applicable additional found)

- key word searches on the LER In the current version of the flooding PRA, industry data applied to develop requirements in 2-2.1.

database are not comprehensive (this appears to be the maintenance contribution to flood frequency is based on flood events because some utilities did not provide any key words documented from 1970

- 2011 in the PIPExp database as described in on their LERs, or the key words provided are not EA-PSA-FLOOD-lE-13-02 Rev. 0, 9ntemal Flood Initiating Event consistent or comprehensive).

Frequencies for the Palisades PRA [38]. This database is also used in As evidenced by the 5750 report, an LER search the latest EPRI guidance for establishing maintenance contribution to using key words is not comprehensive or complete.

flooding (Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Since the 5750 report only covers a subset of the 16 Assessments Revision 3, 2013, Report 30020000079 [37]). The database years of LERs that are being considered for the is compiled by experts and judged to be comprehensive enough to generic prior data, it is probable that additional establish an appropriate prior for Bayes update with Palisades specific applicable LER5 were missed.

data.

Since the population is so small, missing even 1 LER Finding Resolved has an impact on the internal-flood frequency.

The analysis performed could perform a review of all LERs in the INPO LER Database over the period of 1987-2002 to identify potentially missed Internal-flooding LERs (would be very time-intensive) to ensure completeness.

OR A number of utilities calculate the internal-flood frequency based on the EPRI TR-1 01341 report (note a newer report is to be issued imminently). This approach could be used for Palisades.

IFEV-A5-02 Finding DETERMINE the flood-When calculating the internal-flooding generic prior, a This finding is related to the development of the overall plant flood initiating event frequency for capacity factor of 75% was assumed.

frequency and its subsequent use to determine the maintenance each flood scenario group The 75% capacity factor was stated to be assumed contribution to flood frequency.

by using the applicable based on industry operating data from 1987-1995 (as An assumed plant capacity factor was not applied in the current flooding requirements in 2-2.1.

reported in NUREG-CR15750). Since this only covers update. Plant availability to establish the industry prior for maintenance a subset of the years contained in the LER review, induced flooding is based on 3,554 reactor calendar years of experience and since the later years (where capacity factors for from all reactors described in the PlPExp database. The database is industry were higher) are not being included, the described in EA-PSA-FLOOD-IE-13-02 Rev. 0 [38] and EPRI Pipe capacity factor is not reflective of the actual operating Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments history, and appears to be under-estimated. Note: a Page 6 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion IFEV-A5-01 Finding IFEV-A5-02 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.558 RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text DETERMINE the flood-Key-words are used to identify the potential applicable This finding is related to the development of the overall plant flood initiating event frequency for LERs in the INPO Database (5 found). However, as frequency and its subsequent use to determine the maintenance each flood scenario group noted by the additional LERs identified in 5750 (3 contribution to flood frequency.

by using the applicable additional found) - key word searches on the LER In the current version of the flooding PRA, industry data applied to develop requirements in 2-2.1.

database are not comprehensive (this appears to be because some utilities did not provide any key words the maintenance contribution to flood frequency is based on flood events on their LERs, or the key words provided are not documented from 1970 - 2011 in the PIPExp database as described in consistent or comprehensive).

EA-PSA-FLOOD-IE-13-02 Rev. 0, "Intemal Flood Initiating Event Frequencies for the Palisades PRA* [38]. This database is also used in As evidenced by the 5750 report, an LER search the latest EPRI guidance for establishing maintenance contribution to using key words is not comprehensive or complete.

flooding (Pipe Rupture Frequencies for Intemal Flooding Probabilistic Risk Since the 5750 report only covers a subset of the 16 Assessments Revision 3, 2013, Report 30020000079 [37]). The database years of LERs that are being considered for the is compiled by experts and judged to be comprehensive enough to generic prior data, it is probable that additional establish an appropriate prior for Bayes update with Palisades' specific applicable LERs were missed.

data.

Since the population is so small, missing even 1 LER Finding Resolved has an impact on the intemal-flood frequency.

The analysis performed could perform a review of all LERs in the INPO LER Database over the period of 1987-2002 to identify potentially missed Intemal-flooding LERs (would be very time-intensive) to ensure completeness.

OR A number of utilities calculate the intemal-flood frequency based on the EPRI TR-101341 report (note a newer report is to be issued imminently). This approach could be used for Palisades.

DETERMINE the flood-When calculating the intemal-flooding generic prior, a This finding is related to the development of the overall plant flood initiating event frequency for capacity factor of 75% was assumed.

frequency and its subsequent use to determine the maintenance each flood scenario group The 75% capacity factor was stated to be assumed contribution to flood frequency.

by using the applicable based on industry operating data from 1987-1995 (as An assumed plant capacity factor was not applied in the current flooding requirements in 2-2.1.

reported in NUREG-CRl5750). Since this only covers update. Plant availability to establish the industry prior for maintenance a subset of the years contained in the LER review, induced flooding is based on 3,554 reactor calendar years of experience and since the later years (where capacity factors for from all reactors described in the PIPExp database. The database is industry were higher) are not being included, the described in EA-PSA-FLOOD-IE-13-02 Rev. 0 [38] and EPRI Pipe capacity factor is not reflective of the actual operating Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments history, and appears to be under-estimated. Note: a Page 6 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) quick calculation of the capacity factor based on Revision 3, 2013, Report 30020000079 [37].

Table Al.4-4 information calculated a capacity factor Finding Resolved

>80% for the years specified in NUREG-CR/5750).

Use the data available on the NRC website, and calculate the actual capacity factor for the years of interest.

IFEV-A5-03 Finding DETERMINE the flood-The plant-specific data is based on operating history This finding is related to the development of the overall plant flood initiating event frequency for for the life of Palisades. The generic priors is based frequency and its subsequent use to determine the maintenance each flood scenario group on industry data from 1987

- 2002 (including contribution to flood frequency.

by using the applicable Palisades data). Need to provide a justification as to In the current update, Palisades specific flood events were excluded from requirements in 2-2.1.

why the overlap of data is acceptable from a Bayesian updating perspective, the prior data for the Bayesian update as documented in Attachment 2 of EA-PSA-INTFLOOD-l3-06 Volume 2 Rev. 0 [35].

Bayesian updating principles require the priors to be Finding Resolved independent of the update data. Remove the Palisades data from the generic priors, or only use Palisades data since 2002.

IFEV-A5-04 Finding DETERMINE the flood-LER Screening criteria Al.3.1 g appears to be non-This finding is related to the development of the overall plant flood initiating event frequency for conservative. This assumption/screening criteria frequency and its subsequent use to determine the maintenance each flood scenario group states: Leaks in the HPSI system or in the diesel contribution to flood frequency.

by using the applicable generator cooling systems were not considered since In the current version of the flooding PRA, industry data applied to develop requirements in 2-2.1.

these systems would be operating only as a result of the maintenance contribution to flood frequency is based on flood events another event. Since testing and maintenance of documented from 1970

- 2011 in the PIPExp database as described in these systems at power also require the systems to EA-PSA-FLOOD-lE-13-02 Rev. 0, Internal Flood Initiating Event be in operation, events associated with these systems should not be excluded (maintenance events could Frequencies for the Palisades PRA [38]. This database is also used in the latest EPRI guidance for establishing maintenance contribution to and probably are

- the most likely source of potential flooding events associated with these systems).

flooding (Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments Revision 3, 2013, Report 30020000079 [37]). The database Since the frequency of maintenance events is based is compiled by experts and judged to be comprehensive enough to on back calculating from the total frequency -

establish an appropriate prior for Bayes update with Palisades specific screening out these events appears to be non-data.

conservative.

Finding Resolved Dont screen out the events associated with systems that can be/are tested/maintained at power.

IFEV-A7-0l Finding INCLUDE consideration of Palisades calculates the human-induced floods during This finding is related to the development of the overall plant flood human-induced floods maintenance by back-calculating the maintenance-frequency and its subsequent use to determine the maintenance during maintenance through induced floods by taking the overall internal-flood contribution to flood frequency.

frequency and subtracting out passive failures which Page 7 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion IFEV-AS-03 Finding IFEV-AS-04 Finding IFEV-A7-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a{a){3){i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text quick calculation of the capacity factor based on Revision 3, 2013, Report 30020000079 [37].

Table A 1.4-4 information calculated a capacity factor Finding Resolved

>80% for the years specified in NUREG-CRlS7S0).

Use the data available on the NRC website, and calculate the actual capacity factor for the years of interest.

DETERMINE the flood-The plant-specific data is based on operating history This finding is related to the development of the overall plant flood initiating event frequency for for the "life" of Palisades. The generic priors is based frequency and its subsequent use to determine the maintenance each flood scenario group on industry data from 1987 - 2002 (including contribution to flood frequency.

by using the applicable Palisades data). Need to provide a justification as to In the current update, Palisades specific flood events were excluded from requirements in 2-2.1.

why the "overlap" of data is acceptable from a the prior data for the Bayesian update as documented in Attachment 2 of Bayesian updating perspective.

EA-PSA-INTFLOOD-13-06 Volume 2 Rev. 0 [3S].

Bayesian updating principles require the priors to be Finding Resolved "independent" of the update data. Remove the Palisades data from the generic priors, or only use Palisades data since 2002.

DETERMINE the flood-LER Screening criteria A 1.3.1 g appears to be non-This finding is related to the development of the overall plant flood initiating event frequency for conservative. This assumption/screening criteria frequency and its subsequent use to determine the maintenance each flood scenario group states: Leaks in the HPSI system or in the diesel contribution to flood frequency.

by using the applicable generator COOling systems were not considered since In the current version of the flooding PRA, industry data applied to develop requirements in 2-2.1.

these systems would be operating only as a result of the maintenance contribution to flood frequency is based on flood events another event. Since testing and maintenance of documented from 1970 - 2011 in the PIPExp database as described in these systems at power also require the systems to be in operation, events associated with these systems EA-PSA-FLOOD-IE-13-02 Rev. 0, "Internal Flood Initiating Event should not be excluded (maintenance events could -

Frequencies for the Palisades PRA" [38]. This database is also used in and probably are - the most likely source of potential the latest EPRI guidance for establishing maintenance contribution to flooding events associated with these systems).

flooding (Pipe Rupture Frequencies for Intemal Flooding Probabilistic Risk Assessments Revision 3, 2013, Report 30020000079 [37]). The database Since the frequency of maintenance events is based is compiled by experts and judged to be comprehensive enough to on "back calculating" from the total frequency -

establish an appropriate prior for Bayes update with Palisades' specific screening out these events appears to be non-data.

conservative.

Finding Resolved Don't screen out the events associated with systems that can be/are tested/maintained at power.

INCLUDE consideration of Palisades calculates the human-induced floods during This finding is related to the development of the overall plant flood human-induced floods maintenance by "back-calculating" the maintenance-frequency and its subsequent use to determine the maintenance during maintenance through induced floods by taking the overall internal-flood contribution to flood frequency.

frequency and subtracting out *passive failures* which Page 7 of 55 I

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&0 #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) application of generic data.

contribute to the frequency. This number is then The approach described in the findings was not applied in the current further reduced by assuming that only 30% of the flooding update. Maintenance induced flooding frequency is calculated maintenance-induced failures would require at power based on the approach described in EA-PSA-FLOOD-IE-13-02 Rev. 0, based on interviews and operating practices from the Internal Flood Initiating Event Frequencies for the Palisades PRA and early 1990s.

documented in Attachment 2 of EA-PSA-INTFLOOD-1 3-06 Volume 0 Rev.

2 [35]. The total maintenance induced contribution to a specific flood The 70/30 split is based on discussions that occurred event is now added directly to the passive pipe failure frequency for during the IPE days (early 1990s). During that time frame, most sites maintenance practices included a purposes of determining the total initiating event frequency. No credit is taken to reduce the maintenance frequency by an assumed ratio of majority of maintenance being performed during maintenance being performed on-line versus off-line.

outages. However, this philosophy has changed for the industry, and maintenance being performed at Finding Resolved power in order to shorten outage durations is much more common. Therefore, the 70/30 split may no longer be applicable.

Since this split is used to reduce the overall internal-flood frequency, it has a direct impact on the internal-flood frequencies for the various scenarios being induced.

IFQU-A3-01 Finding SCREEN OUT a flood area Because Palisades used a truncation limit of 1 E-09/yr, The current analysis evaluates truncation down to 1 E-1 1 to ensure no if the product of the sum of it is potential that 1 of the 2 flood areas that are flood areas were improperly screened. The truncation study is the frequencies of the flood reported as having a CDF < 1 E-9/yr (FOl

- E sfgrd, documented in the main report of EA-PSA-INTFLOOD-1 3-06 Vol. 3 [36].

scenarios for the area, and and F06

- Aux Bldg) may be artificially screened Finding Resolved the bounding conditional even though there is no positive evidence this criteria core damage probability was met for the zone.

(CCDP) is less than 10-Since the east safeguard room has 6 scenarios 9/reactor year. The associated with it, it could exceed the 1 E-9/yr CDF if bounding CCDP is the each of the scenarios were in the 2E-1 0/yr range.

highest of the CCDP values Based on the CDF summary provided in NB-PSA-lF for the flood scenarios in an Rev. 0, the 2 zones with a CDF <1 E-9/yr are not area.

considered as one of the eleven flood zones defined for the Palisades Plant, so it may have been inappropriately screened.

Lower the truncation limit used during quantification.

IFQU-A6-01 Finding For all human failure events In Section 7.3.2 of EA-PSA-INTFLOOD 03(03),

All FPIE HEPs that have applicability to the flooding model were re in the internal flood Palisades states Human errors developed as a part screened and appropriate shaping factors were applied specifically for scenarios, INCLUDE the of the internal events PRA have been left at their flood initiating events. These events were also screened for operator following scenario-specific existing failure probabilities. This would be reasonable accessibility (Attachment 1 of [36]) to ensure the action could still be impacts on PSFs for control given that plant response to transient and loss of completed when considering water propagation and submergence of Page 8 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion IFQU-A3-01 Finding IFQU-A6-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text application of generic data.

contribute to the frequency. This number is then The approach described in the findings was not applied in the current further reduced by assuming that only 30% of the flooding update. Maintenance induced flooding frequency is calculated maintenance-induced failures would require at power based on the approach described in EA-PSA-FLOOD-IE-13-02 Rev. 0, based on interviews and operating practices from the "Intemal Flood Initiating Event Frequencies for the Palisades PRAA and early 1990's.

documented in Attachment 2 of EA-PSA-INTFLOOD-13-06 Volume 0 Rev.

The 70/30 split is based on discussions that occurred 2 [35]. The total maintenance induced contribution to a specific flood during the IPE days (early 1990's). During that time event is now added directly to the passive pipe failure frequency for purposes of determining the total initiating event frequency. No credit is frame, most sites' maintenance practices included a taken to reduce the maintenance frequency by an assumed ratio of majority of maintenance being performed during maintenance being performed on-line versus off-line.

outages. However, this philosophy has changed for the industry, and maintenance being performed at Finding Resolved power in order to shorten outage durations is much more common. Therefore, the 70/30 split may no longer be applicable.

Since this split is used to reduce the overall intemal-flood frequency, it has a direct impact on the intemal-flood frequencies for the various scenarios being induced.

SCREEN OUT a flood area Because Palisades used a truncation limit of 1 E-09/yr, The current analysis evaluates truncation down to 1 E-11 to ensure no if the product of the sum of it is potential that 1 of the 2 flood areas that are flood areas were improperly screened. The truncation study is the frequencies of the flood reported as having a CDF < 1 E-9/yr (F01 - E sfgrd, documented in the main report of EA-PSA-INTFLOOD-13-06 Vol. 3 [36].

scenarios for the area, and.

and F06 - Aux Bldg) may be artificially Ascreened" Finding Resolved the bounding conditional even though there is no positive evidence this criteria core damage probability was met for the zone.

(CCDP) is less than 10-Since the east safeguard room has 6 scenarios 9/reactor year. The bounding CCDP is the associated with it, it could exceed the 1 E-9/yr CDF if highest of the CCDP values each of the scenarios were in the 2E-1 O/yr range.

for the flood scenarios in an Based on the CDF summary provided in NB-PSA-IF area.

Rev. 0, the 2 zones with a CDF <1 E-9/yr are not considered as one of the eleven flood zones defined for the Palisades Plant, so it may have been inappropriately screened.

Lower the truncation limit used during quantification.

For all human failure events In Section 7.3.2 of EA-PSA-INTFLOOD 03(03),

All FPIE HEPs that have applicability to the flooding model were re-in the intemal flood Palisades states "Human errors developed as a part screened and appropriate shaping factors were applied specifically for scenarios, INCLUDE the of the intemal events PRA have been left at their flood initiating events. These events were also screened for operator following scenario-specific existing failure probabilities. This would be reasonable accessibility (Attachment 1 of [36]) to ensure the action could still be impacts on PSFs for control given that plant response to transient and loss of completed when considering water propagation and submergence of Page 8 of 55 I

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) room and ex-control room offsite power related events should be similar equipment. Actions that could not be performed due to flood or spray from actions as appropriate to the regardless of the exact cause of the initiating event, high energy lines preventing access were set True in the flood model for HRA methodology being However, with the additional complication of a flood, the specific flood initiating event.

used: (a) additional performance shaping factors (PSFs) in the internal Finding Resolved workload and stress (above events PRA may not be as appropriate.

1 As part of that for similar sequences their quantification, Palisades did not change the not caused by internal HEP5 from the internal events HFEs. Therefore, floods) (b) cue availability (c)

Palisades did not address the flood specific impacts effect of flood on mitigation, on the PSF5.

required response, timing, and recovery activities (e.g.,

The flood specific impacts are such that the HEPs carried over are non-conservatively low.

accessibility restrictions, possibility of physical harm)

Revise the quantification of the internal events HEPs (d) flooding-specific job aids to address the impact of the flood on the PSFs.

and training (e.g.,

procedures, training exercises)

IFQU-A7-01 Finding PERFORM internal flood A truncation limit of 1 E-09/yr was used for the Internal The current analysis evaluates truncation down to 1 E-1 1 to ensure no sequence quantification in Flooding analysis. The acceptability of this truncation flood areas were improperly screened and that the requirements of QU-B3 accordance with the limit was not provided, are met. The truncation study is documented in the main report of applicable requirements EA-PSA-INTFLOOD-13-06 Vol. 3 [36].

There is no evidence that this truncation is sufficiently described in 2-2.7 low to meet the requirements of QU-B3 Finding Resolved (demonstrates that the overall internal flood model results converge and no significant accident sequences are inadvertently eliminated.)

Lower the truncation limit until convergence is obtained.

IFQU-A9-01 Finding INCLUDE, in the A specific discussion of jet impingement and pipe Added discussion to Section A3.5 in Attachment 3 of EA-PSA-INTFLOOD quantification, both the whips was not identified.

13-06 Volume 1 Rev. 0 [35] regarding how pipe whip and jet impingement direct effects of the flood were evaluated during the ISI walkdowns; as the RI-ISI report provides the Consideration of jet impingement and pipe whips (as basis for the indirect effects on equipment for each flood or spray initiator.

(e.g., loss of cooling from a service water train due to an appropriate) are a requirement of the standard for this The RI-ISI walkdowns were not limited in time or scope due to escort element.

associated pipe rupture) and issues. This statement was only applicable to the 2008 update walkdown.

indirect effects such as Provide a discussion of how jet impingements and Detailed walkdowns of all flood areas were documented in the RI-ISI submergence, jet pipe whips were considered and handled. The indirect effects report EA-PSA-lSl-00-INDIRECT Rev. 0 [39].

impingement, and pipe whip, Internal Flooding Analysis Report referenced Additional walkdown documentation and clarification was added to as applicable.

walkdowns performed for the IPE. The scope of these of EA-PSA-INTFLOOD-13-06 Vol. 1 Rev. 0 [34] to walkdown was limited as a result of time constraints demonstrate additional walkdowns were performed both before and after placed on the walkdown team by the authorized team Page 9 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion IFQU-A7-01 Finding IFQU-A9-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text room and ex-control room offsite power related events should be similar equipment. Actions that could not be performed due to flood or spray from actions as appropriate to the regardless of the exact cause of the initiating event.

high energy lines preventing access were set 'True' in the flood model for HRA methodology being However, with the additional complication of a flood, the specific flood initiating event.

used: (a) additional performance shaping factors (PSFs) in the internal Finding Resolved workload and stress (above events PRA may not be as appropriate. Q As part of that for similar sequences their quantification, Palisades did not change the not caused by internal HEPs from the internal events HFEs. Therefore, floods) (b) cue availability (c)

Palisades did not address the flood specific impacts effect of flood on mitigation, on the PSFs.

required response, timing, The flood specific impacts are such that the HEPs and recovery activities (e.g.,

carried over are non-conservatively low.

accessibility restrictions, possibility of phYSical harm)

Revise the quantification of the internal events HEPs (d) flooding-specific job aids to address the impact of the flood on the PSFs.

and training (e.g.,

procedures, training exercises)

PERFORM internal flood A truncation limit of 1 E-09/yr was used for the Internal The current analysis evaluates truncation down to 1 E-11 to ensure no sequence quantification in Flooding analysis. The acceptability of this truncation flood areas were improperly screened and that the requirements of QU-B3 accordance with the limit was not provided.

are met. The truncation study is documented in the main report of applicable requirements There is no evidence that this truncation is suffiCiently EA-PSA-INTFLOOD-13-06 Vol. 3 [36].

described in 2-2.7 low to meet the requirements of QU-S3 Finding Resolved (demonstrates that the overall internal flood model results converge and no significant accident sequences are inadvertently eliminated.)

Lower the truncation limit until convergence is obtained.

INCLUDE, in the A specific discussion of jet impingement and pipe Added discussion to Section A3.S in Attachment 3 of EA-PSA-INTFLOOD-quantification, both the whips was not identified.

13-06 Volume 1 Rev. 0 [35] regarding how pipe whip and jet impingement direct effects of the flood Consideration of jet impingement and pipe whips (as were evaluated during the lSI walkdowns; as the RI-ISI report provides the (e.g., loss of cooling from a appropriate) are a requirement of the standard for this basis for the indirect effects on equipment for each flood or spray initiator.

service water train due to an The RI-ISI walkdowns were not limited in time or scope due to escort associated pipe rupture) and element.

issues. This statement was only applicable to the 2008 update walkdown.

indirect effects such as Provide a discussion of how jet impingements and Detailed walkdowns of all flood areas were documented in the RI-ISI submergence, jet pipe whips were considered and handled. The indirect effects report EA-PSA-ISI-OO-INDIRECT Rev. 0 [39].

impingement, and pipe whip, Internal Flooding Analysis Report referenced Additional walkdown documentation and clarification was added to as applicable.

walkdowns performed for the I PE. The scope of these of EA-PSA-INTFLOOD-13-06 Vol. 1 Rev. 0 [34] to walkdown was limited as a result of time constraints demonstrate additional walkdowns were performed both before and after placed on the walkdown team by the authorized team Page 9 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) escort. Palisades indicated that they had performed a the 2008 walkdown in which the escort had limited time.

more recent complete walkdown, but that walkdown Finding Resolved was not referenced in the Internal Flooding Analysis Report. The consideration of jet impingement and pipe whip is qualitatively and semi-quantitatively discussed in the walkdown notes for the more recent walkdown. It Palisades wants to credit the more recent walkdown, they need to reference it in the Internal Flooding Analysis Report.

IFSN-A15-01 Finding For each defined flood area The Heater Drain pump suction tank T-5 has Tank T-5 is located just above the 590 elevation of the Turbine Building and each flood source, insufficient capacity to flood the room. Tank T-60 NA north side. The EDG rooms are located adjacent to the Turbine Building IDENTIFY the propagation Dirty Waste Drain Tank RI-In-Service Inspection (ISI) north hallway 590 elevation and are protected by water tight doors in this path from the flood source does not evaluate tanks, only pipes. However, this area. Therefore, flooding originating in the Turbine Building cannot area to its area of tank has insufficient volume to flood area to any propagate to the EDG rooms. This basis was added as a clarification to accumulation.

significant height. HBD-13-3 Misc West Drain Tank the walkdown documentation in Attachment 5 of T89A/B From Spool To Condensate Storage Tank EA-PSA-INTFLOOD-13-06 Vol. 1 Rev. 0 [34].

Water 20. The documentation states that it is assumed that there is insufficient volume to flood to Finding Resolved level of EDG

- but the justification/basis for this assumption is not provided.

Need to verify that the basis for the assumption is valid or an additional EDG failure mode could be missed.

Provide basis for determining or assuming insufficient volume.

IFSN-A17-01 Finding CONDUCT a plant The scope of the walkdown was limited to the Detailed walkdowns of all flood areas were documented in the RI-ISI walkdown(s) to verify the identified areas as a result of time constraints placed indirect effects report EA-PSA-ISI-00-INDIRECT Rev. 0 [39]. Additional accuracy of information on the walkdown team by the authorized team escort.

walkdown documentation and clarification was added to Attachment 5 of obtained from plant His limited availability resulted in the walkdown team EA-PSA-INTFLOOD-13-06 Vol. 1 Rev. 0 [34] to demonstrate additional information sources and to prioritizing the areas reviewed.

walkdowns were performed both before and after the 2008 walkdown in obtain or verify (a) SSCs which the escort had limited time.

Because of the limited walkdown time, some rooms located within each defined were not walked down, therefore the pipe lengths for The current methodology does not rely on the RI-ISI derived pipe failure flood area (b) these rooms were not identified. This results in the frequency data. All pipe failure frequencies were developed in flood/spray/other applicable mitigative features of the frequency for the rooms relying on Rl-lSl data accordance with latest EPRI methodology [37]. All pipe lengths were instead of being able to use the best available data for obtained from plant isometric drawings.

SSCs located within each pipe/component failure rates.

defined flood area (e.g.,

Finding Resolved drains, shields, etc.) (c)

The Internal Flooding Analysis Report referenced pathways that could lead to walkdowns performed for the IPE. The scope of these Page 10 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion IFSN-A 15-01 Finding I FSN-A 17-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text escort. Palisades indicated that they had performed a the 2008 walkdown in which the escort had limited time.

more recent complete walkdown, but that walkdown Finding Resolved was not referenced in the Internal Flooding Analysis Report. The consideration of jet impingement and pipe whip is qualitatively and semi-quantitatively discussed in the walkdown notes for the more recent walkdown. If Palisades wants to credit the more recent walkdown, they need to reference it in the Internal Flooding Analysis Report.

For each defined flood area The Heater Drain pump suction tank T -5 has Tank T-5 is located just above the 590' elevation of the Turbine Building and each flood source, insufficient capacity to flood the room. Tank T-60 NA north side. The EDG rooms are located adjacent to the Turbine Building IDENTIFY the propagation Dirty Waste Drain Tank RI-In-Service Inspection (lSI) north hallway 590' elevation and are protected by water tight doors in this path from the flood source does not evaluate tanks, only pipes. However, this area. Therefore, flooding originating in the Turbine Building cannot area to its area of tank has insufficient volume to flood area to any propagate to the EDG rooms. This basis was added as a clarification to accumulation.

Significant height. HBD-13-3 Misc West Drain Tank the walkdown documentation in Attachment 5 of T89AIB From Spool To Condensate Storage Tank EA-PSA-INTFLOOD-13-06 Vol. 1 Rev. 0 [34].

Water 20. The documentation states that it is Finding Resolved assumed that there is insufficient volume to flood to level of EDG - but the justificationlbasis for this assumption is not provided.

Need to verify that the basis for the assumption is valid or an additional EDG failure mode could be missed.

Provide basis for determining or assuming insufficient volume.

CONDUCT a plant The scope of the walkdown was limited to the Detailed walkdowns of all flood areas were documented in the RI-ISI walkdown(s) to verify the identified areas as a result of time constraints placed indirect effects report EA-PSA-ISI-OQ-INDIRECT Rev. 0 [39]. Additional accuracy of information on the walkdown team by the authorized team escort.

walkdown documentation and clarification was added to Attachment 5 of obtained from plant His limited availability resulted in the walkdown team EA-PSA-INTFLOOD-13-06 Vol. 1 Rev. 0 [34] to demonstrate additional information sources and to prioritizing the areas reviewed.

walkdowns were performed both before and after the 2008 walkdown in obtain or verify (a) SSCs Because of the limited walkdown time, some rooms which the escort had limited time.

located within each defined were not walked down, therefore the pipe lengths for The current methodology does not rely on the RI-ISI derived pipe failure flood area (b) these rooms were not identified. This results in the frequency data. All pipe failure frequencies were developed in flood/spray/other applicable nfrequency* for the rooms relying on RI-ISI data accordance with latest EPRI methodology [37]. All pipe lengths were mitigative features of the instead of being able to use the best available data for obtained from plant isometric drawings.

SSCs located within each defined flood area (e.g.,

pipe/component failure rates.

Finding Resolved drains, shields, etc.) (c)

The Internal Flooding Analysis Report referenced pathways that could lead to walkdowns performed for the IPE. The scope of these Page 10 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions Requement)

Andingor 1.200 Finding Description (summary discussion)

Disposition transport to the flood area walkdown was limited as a result of time constraints placed on the walkdown team by the authorized team escort. Palisades indicated that they had performed a more recent complete walkdown, but that walkdown was not referenced in the Internal Flooding Analysis Report. If Palisades wants to credit the more recent walkdown, they need to reference it in the Internal Flooding Analysis Report.

IFSN-A3-01 Finding For each defined flood area Those automatic or operator responses that have the Palisades has developed a flood mitigation abnormal operating procedure and each flood source, ability to terminate or contain the flood propagation for AOP-39 [40] which defines operator actions for flood mitigation in all 11 IDENTIFY those automatic each defined flood area and flood source were not PRA defined flood areas.

or operator responses that identified.

Finding Resolved have the ability to terminate or contain the flood propagation.

Required by SR.

Identify and document the automatic and operator responses that do have the ability to terminate or contain the flood propagation for each defined flood area and source.

IFSN-A6-01 Finding For the SSCs identified in Spray effects from chilled water systems pipe failures Further documentation for eliminating the chilled water system as a flood IFSN-A5, IDENTIFY the dont seem to be addressed. The basis for elimination source is based on plant drawings of the rooms transgressed and design susceptibility of each SSC in as a spray consideration is only documented for 1 information of the Bus 1 D room cooler. This documentation was added to a flood area to flood-induced room

- not for all rooms transgressed. of EA-PSA-INTFLOOD-13-06 Vol. 2 [35].

failure mechanisms.

INCLUDE failure by This requirement states that the susceptibility of each Finding Resolved SSC in a flood area to flood-induced failures submergence and spray in the identification process.

mechanisms by either submergence or spray are EITHER: a) ASSESS included.

qualitatively the impact of Address the potential spray effects from chilled water flood-induced mechanisms system pipe failures in all zones transgressed.

that are not formally addressed (e.g., using the mechanisms listed under Capability Category III of this requirement), by using conservative assumptions; OR b) NOTE that these mechanisms are not Page 11 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion I FSN-A3-01 Finding I FSN-A6-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text transport to the flood area walkdown was limited as a result of time constraints placed on the walkdown team by the authorized team escort. Palisades indicated that they had performed a more recent complete walkdown, but that walkdown was not referenced in the Internal Flooding Analysis Report. If Palisades wants to credit the more recent walkdown, they need to reference it in the Internal Flooding Analysis Report.

For each defined flood area Those automatic or operator responses that have the Palisades has developed a flood mitigation abnormal operating procedure and each flood source, ability to terminate or contain the flood propagation for AOP-39 [40] which defines operator actions for flood mitigation in all 11 IDENTIFY those automatic each defined flood area and flood source were not PRA defined flood areas.

or operator responses that identified.

Finding Resolved have the ability to terminate or contain the flood propagation.

Required by SA.

Identify and document the automatic and operator responses that do have the ability to terminate or contain the flood propagation for each defined flood area and source.

For the SSCs identified in Spray effects from chilled water systems pipe failures Further documentation for eliminating the chilled water system as a flood IFSN-A5, IDENTIFY the don't seem to be addressed. The basis for elimination source is based on plant drawings of the rooms transgressed and design susceptibility of each SSC in as a spray consideration is only documented for 1 information of the Bus 1 D room cooler. This documentation was added to a flood area to flood-induced room - not for all rooms transgressed. of EA-PSA-INTFLOOD-13-06 Vol. 2 [35].

failure mechanisms.

This requirement states that the susceptibility of each Finding Resolved INCLUDE failure by SSC in a flood area to flood-induced failures submergence and spray in mechanisms by either submergence or spray are the identification process.

included.

EITHER: a) ASSESS qualitatively the impact of Address the potential spray effects from chilled water flood-induced mechanisms system pipe failures in all zones transgressed.

that are not formally addressed (e.g., using the mechanisms listed under Capability Category III of this requirement), by using conservative assumptions; OR b) NOTE that these mechanisms are not

-- L....

Page 11 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 5055a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) included in the scope of the evaluation.

IFSO-A4-01 Finding For each potential source of Palisades did not explicitly identify and characterize Human induced flood events are characterized for each flood area flooding water, IDENTIFY human induced flooding events for each flood area.

initiating event as part of the maintenance induced flooding frequency the flooding mechanisms Instead, Palisades chose to characterize the human-development. Maintenance induced flood frequency in each flood area is that would result in a fluid induced flooding events by setting a generic element system specific to characterize the flood mechanism.

releaser. INCLUDE: (a) and then back-calculating a frequency without actually failure modes of delineating what the human induced event was.

The approach to this is described in reports EA-PSA-FLOOD-lE-13-02 components such as pipes, Rev. 0 [38] and Attachment 2 of EA-PSA-INTFLOOD-13-06 Volume 2 Without a reasonable characterization of the specific

[35].

tanks, gaskets, expansion human induced flooding events it is difficult to joints, fittings, seals, etc. (b) understand their full impact on the results or address Finding Resolved human-induced them should they be found to be significant mechanisms that could lead contributors.

to overfilling tanks, diversion of flow through openings Palisades should either more fully characterize the created to perform human induced flooding events or they should be maintenance; inadvertent explicitly called out as assumptions so that they can actuation of fire suppression be assessed for applications affecting internal system (c) other events flooding.

resulting in a release into the flood area IFSN-A1 2-01 Suggestion For each defined flood area Although the screening of rooms appears to be The flood area selection was re-validated per the screening criteria of (Suggestion) and each flood source, reasonable, it is not clear what criteria from the Entergy Internal Flooding Guidelines procedure EN-NE-G-012 Section IDENTIFY the propagation Standard was used for the various flood areas 5.3.3 [41]. This procedure was added as a reference document to the path from the flood source screened. Because of the multiple screening criteria flood area development documentation in Attachment 1 of area to its area of available, specifying the criteria applied would be EA-PSA-INTFLOOD-13-06 Vol. 1 [34].

accumulation.

beneficial to ensure no zones were inappropriately screened.

Suggestion Resolved This may result in additional zones being able to be screened, and would ensure the zones already screened were done in accordance with the Standards requirements.

Specify the criteria from the Standard that was applied to screen the various zones from further consideration.

IFSN-A8-01 Suggestion IDENTIFY inter-area Inter-area propagation through the normal flow path Further documentation for eliminating the chilled water system as a flood (Suggestion) propagation through the from one area to another via drain lines were source is based on plant drawings of the rooms transgressed and design normal flow path from one addressed by the GOTHIC runs. Areas connected via information of the Bus 1 D room cooler. This documentation was added to Page 12 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion I FSO-A4-01 Finding IFSN-A12-01 Suggestion (Suggestion)

I FSN-A8-01 Suggestion (Suggestion)

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text included in the scope of the evaluation.

For each potential source of Palisades did not explicitly identify and characterize Human induced flood events are characterized for each flood area flooding water, IDENTIFY human induced flooding events for each flood area.

initiating event as part of the maintenance induced flooding frequency the flooding mechanisms Instead, Palisades chose to characterize the human-development. Maintenance induced flood frequency in each flood area is that would result in a fluid induced flooding events by setting a generic element system specific to characterize the flood mechanism.

releaser. INCLUDE: (a) and then back-calculating a frequency without actually The approach to this is described in reports EA-PSA-FLOOD-IE-13-02 failure modes of delineating what the human induced event was.

Rev. 0 [38] and Attachment 2 of EA-PSA-INTFLOOD-13-06 Volume 2 components such as pipes, Without a reasonable characterization of the specific

[35].

tanks, gaskets, expansion human induced flooding events it is difficult to joints, fittings, seals, etc. (b) understand their full impact on the results or address Finding Resolved human-induced mechanisms that could lead them should they be found to be significant to overfilling tanks, diversion contributors.

of flow through openings Palisades should either more fully characterize the created to perform human induced flooding events or they should be maintenance; inadvertent explicitly called out as assumptions so that they can actuation of fire suppression be assessed for applications affecting intemal system (c) other events flooding.

resulting in a release into the flood area For each defined flood area Although the screening of rooms appears to be The flood area selection was re-validated per the screening criteria of and each flood source, reasonable, it is not clear what criteria from the Entergy Intemal Flooding Guidelines procedure EN-NE-G-012 Section IDENTIFY the propagation Standard was used for the various flood areas 5.3.3 [41]. This procedure was added as a reference document to the path from the flood source screened. Because of the multiple screening criteria flood area development documentation in Attachment 1 of area to its area of available, specifying the criteria applied would be EA-PSA-INTFLOOD-13-06 Vol. 1[34].

accumulation.

beneficial to ensure no zones were inappropriately Suggestion Resolved screened.

This may result in additional zones being able to be screened, and would ensure the zones already screened were done in accordance with the Standard's requirements.

Specify the criteria from the Standard that was applied to screen the various zones from further consideration.

IDENTIFY inter-area Inter-area propagation through the normal flow path Further documentation for eliminating the chilled water system as a flood propagation through the from one area to another via drain lines were source is based on plant drawings of the rooms transgressed and design normal flow path from one addressed by the GOTHIC runs. Areas connected via information of the Bus 1 D room cooler. This documentation was added to Page 12 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) area to another via drain backflow through drain lines involving failed check of EA-PSA-INTFLOOD-13-06 Vol. 2 [35] and included lines; and areas connected valves, pipe and cable penetrations (including cable discussion of the potential flow rate.

via backflow through drain trays), doors, stairwells, hatchways. There didnt Suggestion Resolved lines involving failed check appear to be any evaluation of the chilled water valves, pipe and cable system flow rates/propagation from pumps through penetrations (including cable piping through the room coolers.

trays), doors, stairwells, Add documentation on the chilled water system and hatchways, and HVAC the flow rates/propagation potential.

ducts. INCLUDE potential for structural failure (e.g., of doors or walls) due to flooding loads.

IFSO-A1-01 Suggestion For each flood area, The chilled water system was identified as being in Further documentation for eliminating the chilled water system as a flood (Suggestion)

IDENTIFY the potential the Bus 1 D, Cable Spreading and Electrical source is based on plant drawings of the rooms transgressed and design sources of flooding [Note Equipment rooms, but was eliminated from information of the Bus 1 D room cooler. This documentation was added to (1)]. INCLUDE: (a) consideration as a flood source because it has of EA-PSA-INTFLOOD-13-06 Vol. 2 [35] and included equipment (e.g., piping, insufficient volume to flood these areas (Ref. [4],

discussion of the potential flow rate.

valves, pumps) located in Appendix A, Final List Of Potential Suggestion Resolved the area that are connected Hazards/Postulated Effects In The Bus 1 D Room).

to fluid systems (e.g.,

Table A2.8-3c: Sources Not Considered for the circulating water system, Internal Flood analysis update, from the plant service water system, walkdown supporting the IPE (Ref. [6]). The basis for component cooling water the insufficient capacity cannot be found.

system, feedwater system, Since the chilled water was eliminated in the original condensate and steam IPE, need to verity and document that the elimination systems) (b) plant internal criteria used is still valid, especially since there may sources of flooding (e.g.,

have been chillers installed in the plant that use tanks or pools) located in Chilled Water since the IPE was performed.

the flood area (c) plant external sources of flooding Provide the criteria/basis for how it was determined (e.g., reservoirs or rivers) that the Chilled Water system could be eliminated that are connected to the from flooding impacts, and ensure the basis is still area through some system valid.

or structure (d) in-leakage from other flood areas (e.g.,

back flow through drains, doorways, etc.)

AS-A2-01 Finding For each modeled initiating The event trees specify the required key safety Notebook NB-PSA-SS, Palisades Safe and Stable States [23] was event, IDENTIFY the key functions needed to mitigate the initiating event of developed to evaluate and document the non-core damage end states for safety functions that are interest, but the mission time is specified as 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

all event trees. Generalized flow charts were developed to capture all of Page 13 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion IFSO-A1-01 Suggestion (Suggestion)

AS-A2-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a{a){3){i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text area to another via drain backflow through drain lines involving failed check of EA-PSA-INTFLOOD-13-06 Vol. 2 [35] and included lines; and areas connected valves, pipe and cable penetrations (including cable discussion of the potential flow rate.

via backflow through drain trays), doors, stairwells, hatchways. There didn't Suggestion Resolved lines involving failed check appear to be any evaluation of the chilled water valves, pipe and cable system flow rates/propagation from pumps through penetrations (including cable piping through the room coolers.

trays), doors, stairwells, Add documentation on the chilled water system and hatchways, and HVAC ducts. INCLUDE potential the flow rates/propagation potential.

for structural failure (e.g., of doors or walls) due to flooding loads.

For each flood area, The chilled water system was identified as being in Further documentation for eliminating the chilled water system as a flood IDENTIFY the potential the Bus 1 D, Cable Spreading and Electrical source is based on plant drawings of the rooms transgressed and design sources of flooding [Note Equipment rooms, but was eliminated from information of the Bus 1 D room cooler. This documentation was added to (1)). INCLUDE: (a) consideration as a flood source because it has of EA-PSA-INTFLOOD-13-06 Vol. 2 [35] and included equipment (e.g., piping, insufficient volume to flood these areas (Ref. [4],

discussion of the potential flow rate.

valves, pumps) located in Appendix A, "Final List Of Potential Suggestion Resolved the area that are connected Hazards/Postulated Effects In The Bus 1 D Room").

to fluid systems (e.g.,

Table A2.8-3c: Sources Not Considered for the circulating water system, Internal Flood analysis update, from the plant service water system, walkdown supporting the IPE (Ref. [6]). The basis for component cooling water the insufficient capacity cannot be found.

system, feedwater system, Since the chilled water was eliminated in the original condensate and steam systems) (b) plant internal IPE, need to verify and document that the elimination sources of flooding (e.g.,

criteria used is still valid, especially since there may tanks or pools) located in have been chillers installed in the plant that use the flood area (c) plant Chilled Water since the IPE was performed.

external sources of flooding Provide the criterialbasis for how it was determined (e.g., reservoirs or rivers) that the Chilled Water system could be eliminated that are connected to the from flooding impacts, and ensure the basis is still area through some system valid.

or structure (d) in-leakage from other flood areas (e.g.,

back flow through drains, doorways, etc.)

For each modeled initiating The event trees specify the required key safety Notebook NB-PSA-SS, "Palisades Safe and Stable States" [23] was event, IDENTIFY the key functions needed to mitigate the initiating event of developed to evaluate and document the non-core damage end states for safety functions that are interest, but the mission time is specified as 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

all event trees. Generalized flow charts were developed to capture all of Page 13 of 55 I

I I

I I

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 201 0 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) necessary to reach a safe, Need to ensure all end states are safe, stable states the non-core damage sequences based on the general transient/main stable state and prevent at the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time, or extend the mission steam line break, loss of offsite power, loss of cooling accident, very small core damage.

time until a safe, stable end state is met for each break loss of coolant accident (consequential LOCA) and steam generator accident sequence.

tube rupture event trees. Event tree headings were translated in the flow charts to decision boxes allowing a path to be followed to reach the OK There is not sufficient documentation that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is end states.

appropriate to ensure that all accident sequences reach a safe, stable end state. Also, not all end states Finding Resolved.

specified on the event trees may be correct.

AS-A3-01 Finding For each modeled initiating The documentation associated with the event trees The success criteria notebook, NB-PSA-ETSC [1 1] was revised to ensure event, using the success does not always match the current event tree logic, all event tree headings match the headings described in the notebook criteria defined for each key For some of the event tree nodes there appears to be documentation. Section 4.10 was added to the notebook to describe the operation of auxiliary feedwater pump P-8B after battery depletion safety function (in a documentation mis-match. For example: Section 5.9 heading. Section 5.9 was corrected to agree with the number of required accordance with SR SC-A3),

of NP-PSA-ETSC, rOl states that 3 of 3 charging charging pumps as described in the success criteria in Section 5.9.4 for IDENTIFY the systems that can be used to mitigate the pumps are required for a VSBLOCA, but the success the CHRG-FT event tree heading.

criteria for the event tree top logic (CHRG-FT) states initiator.

the success criteria is 2 of 3 charging pumps.

Finding Resolved.

AS-Al 0-01 Finding In constructing the accident Although the event trees include operator actions A copy of the Human Error Probability (HEP) Post-Initiator Calculation (P sequence models, required for success of key safety functions, the IC) and associated Post-Initiator Operator Action Questionnaire (P-IOAQ)

INCLUDE, for each modeled documented actions do not include verification that were provided currently SRO licensed on-shift Operations Department initiating event, sufficient the operator actions, as evaluated, are bounding for personnel and Training Department personnel for use in validating HEP detail that significant all event tree nodes where the operator action is information accuracy. A document outlining expectations for the review differences in requirements

applied, was also provided.

on systems and required The CAT II requirement to capture and provide HEP5 were assigned to the five Operations Department Operating Crews operator interactions (e.g.,

sufficient detail for significant differences in (10 per crew) for review. Their reviews included ensuring indications, systems initiations or valve alignment) are captured.

requirements associated with systems and/or procedure selection and use, and activity performance man-power and Where diverse systems operator responses is not performed. For example, timing are correct. Training personnel reviews included ensuring the event tree node PORV-FT appears in multiple procedure selection and use were consistent with current training and/or operator actions event trees including Main Steam Line Break (MSLB),

expectations, and the training type and frequency are accurate.

provide a similar function, if SGTR, LOBUS1A, PCP-SBLOCA, LOOP, but the choosing one over another Operator action is based on timing for Loss of Main Operator comments were reviewed and proposed resolutions forwarded to the comment initiator for further comment or acceptance. Initiator changes the requirements Feedwater (LOMFW). There is no differentiation comment acceptance was documented by their initialing the HEP for operator intervention or between the timing for any of the other initiators, and Validation form.

the need for other systems, it does not appear that the LOFW initiating event is MODEL each separately.

the bounding event for this operator action.

Training comments were similarly dispositioned and documented. All validations have been completed and the documents attached to Volume I The operator actions as currently evaluated need to of the human reliability analysis notebook (Appendix F of [8]).

be reviewed to ensure they are bounding for all scenarios where they are credited. To meet the CAT Finding Resolved.

II requirement,_timing_differences_(and_potentially Page 14 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion AS-A3-01 Finding AS-A10-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a{a){3){i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text necessary to reach a safe, Need to ensure all end states are safe, stable states the non-core damage sequences based on the general transient/main stable state and prevent at the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time, or extend the mission steam line break, loss of offsite power, loss of cooling accident, very small core damage.

time until a safe, stable end state is met for each break loss of coolant accident (consequential LOCA) and steam generator accident sequence.

tube rupture event trees. Event tree headings were translated in the flow There is not sufficient documentation that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is charts to decision boxes allowing a path to be followed to reach the "OK" appropriate to ensure that all accident sequences end states.

reach a safe, stable end state. Also, not all end states Finding Resolved.

specified on the event trees may be correct.

For each modeled initiating The documentation associated with the event trees The success criteria notebook, NB-PSA-ETSC [11] was revised to ensure event, using the success does not always match the current event tree logic.

all event tree headings match the headings described in the notebook criteria defined for each key For some of the event tree nodes there appears to be documentation. Section 4.10 was added to the notebook to describe the safety function (in operation of auxiliary feedwater pump P-8B after battery depletion accordance with SR SC-A3),

a documentation mis-match. For example: Section 5.9 heading. Section 5.9 was corrected to agree with the number of required IDENTIFY the systems that of NP-PSA-ETSC, r01 states that 3 of 3 charging charging pumps as described in the success criteria in Section 5.9.4 for can be used to mitigate the pumps are required for a VSBLOCA, but the success the CHRG-FT event tree heading.

initiator.

criteria for the event tree top logic (CHRG-FT) states the success criteria is 2 of 3 charging pumps.

Finding Resolved.

In constructing the accident Although the event trees include operator actions A copy of the Human Error Probability (HEP) Post-Initiator Calculation (P-sequence models, required for success of key safety functions, the IC) and associated Post-Initiator Operator Action Questionnaire (P-IOAQ)

INCLUDE, for each modeled documented actions do not include verification that were provided currently SRO licensed on-shift Operations Department initiating event, sufficient the operator actions, as evaluated, are "bounding" for personnel and Training Department personnel for use in validating HEP detail that significant all event tree nodes where the operator action is information accuracy. A document outlining expectations for the review differences in requirements applied.

was also provided.

on systems and required The CAT II requirement to capture and provide HEPs were assigned to the five Operations Department Operating Crews operator interactions (e.g.,

systems initiations or valve sufficient detail for significant differences in

(-10 per crew) for review. Their reviews included ensuring indications, alignment) are captured.

requirements associated with systems and/or procedure selection and use, and activity performance man-power and Where diverse systems operator responses is not performed. For example, timing are correct. Training personnel reviews included ensuring and/or operator actions the event tree node PORV-FT appears in multiple procedure selection and use were consistent with current training provide a similar function, if event trees including Main Steam Line Break (MSLB),

expectations, and the training type and frequency are accurate.

SGTR, LOBUS1A, PCP-SBLOCA, LOOP, but the choosing one over another Operator action is based on timing for Loss of Main Operator comments were reviewed and proposed resolutions forwarded to changes the requirements Feedwater (LOMFW). There is no differentiation the comment initiator for further comment or acceptance. Initiator for operator intervention or between the timing for any of the other initiators, and comment acceptance was documented by their initialing the HEP the need for other systems, it does not appear that the LOFW initiating event is Validation form.

MODEL each separately.

the bounding event for this operator action.

Training comments were similarly dispositioned and documented. All The operator actions as currently evaluated need to validations have been completed and the documents attached to Volume I be reviewed to ensure they are "bounding" for all of the human reliability analysis notebook (Appendix F of [8]).

scenarios where they are credited. To meet the CAT Finding Resolved.

II requirement, timing differences (and potentially Page 14 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) stress levels, etc.) need to be addressed for each accident sequence where the operator actions are credited.

AS-C2-01 Finding DOCUMENT the processes There are some event trees that are not well Added Table 3.0-1 and supporting discussion to Section 3.0 of the event used to develop accident documented in the Accident Sequence or Initiating tree and success criteria notebook NB-PSA-ETSC [1 1] to clarify that all sequences and treat Event notebooks. No documentation associated with transient initiators, including controlled manual shutdown are applicable dependencies in accident the success criteria could be found for the Controlled to the general transient event tree and its associated event tree headings sequences, including the Manual Shutdown Event tree. Additionally, there are and success criteria. The table and associated discussion also describes inputs, methods, and results multiple additional event trees (referred to as Special logical operators that are set to True in the event tree rules file to events); (d) the operator lnitiatorsv) in the SAPHIRE program that are not establish the appropriate boundary conditions for each support system actions reflected in the event explicitly discussed within the accident sequence transient event. This discussion justifies the grouping of these initiators as trees, and the sequence-documentation. A discussion of how an FMEA was applicable to the transient event tree.

specific timing and performed to identify plant-specific system initiators is The result of the completed FMEA for all Palisades systems was dependencies that are included in the Initiating Event notebook, however, the traceable to the HRA for FMEA provided in the report is an example FMEA, developed into the Support System to Front-Line System Dependency Matrix and Support System to Support System Dependency Matrix. This these actions; (e) the but the actual FMEA performed is not included or was clarified in Section 2.2 of the initiating event notebook, NB-PSA-IE interface of the accident referenced. No discussion could be found that sequence models with plant identifies how the final support system initiators were

[10]. The final FMEA results were added as Attachment 12 to the initiating events notebook.

damage states; (f) [when identified, how they are grouped, or how the event sequences are modeled tree branches were defined.

Finding Resolved.

using a single top event fault Since these event trees appear to use the same tree] the manner in which branches as other event tree, their grouping needs the requirements for to be discussed, including the appropriateness of accident sequence analysis using the same event tree nodes for the event trees.

have been satisfied.

Without a discussion of the event trees and nodes associated with the support system initiators, there is no documentation that the key safety functions or success criteria defined is appropriate and adequate for them.

Although the Controlled Manual Shutdown is listed in the table in Attachment 3, there is no mention of it in the discussion in Section 3. A discussion of the event needs to be included in Section 3 similar to how the other transients are described. Also need to include the actual FMEA in the documentation or provide a valid reference for it.

AS-C2-02 Suggestion DOCUMENT the processes While it is obvious that the Flag files exist, the A summary discussion of the event tree rules (Flag) file was added to used to develop accident development and review of the Flag file used in Section 3.0 of the event trees and success criteria notebook NB-PSA sequences and treat SAPH IRE is not included in the accident sequence ETSC [1 1]. In addition, Table 3.0-1 was added which lists all of the Page 15 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion AS-C2-01 Finding AS-C2-02 Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a{a){3){i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text stress levels, etc.) need to be addressed for each accident sequence where the operator actions are credited.

DOCUMENT the processes There are some event trees that are not well Added Table 3.0-1 and supporting discussion to Section 3.0 of the event used to develop accident documented in the Accident Sequence or Initiating tree and success criteria notebook NB-PSA-ETSC [11] to clarify that all sequences and treat Event notebooks. No documentation associated with transient initiators, including 'controlled manual shutdown' are applicable dependencies in accident the success criteria could be found for the Controlled to the general transient event tree and its associated event tree headings sequences, including the Manual Shutdown Event tree. Additionally, there are and success criteria. The table and associated discussion also describes inputs, methods, and results multiple additional event trees (referred to as *Special logical operators that are set to 'True' in the event tree rules file to events); (d) the operator Initiators*) in the SAPHIRE program that are not establish the appropriate boundary conditions for each support system actions reflected in the event explicitly discussed within the accident sequence transient event. This discussion justifies the grouping of these initiators as trees, and the sequence-documentation. A discussion of how an FMEA was applicable to the transient event tree.

specific timing and performed to identify plant-specific system initiators is The result of the completed FMEA for all Palisades systems was dependencies that are included in the Initiating Event notebook, however, the traceable to the H RA for FMEA provided in the report is an "example FMEA",

developed into the 'Support System to Front-Line System Dependency these actions; (e) the but the actual FMEA performed is not included or Matrix' and 'Support System to Support System Dependency Matrix'. This interface of the accident referenced. No discussion could be found that was clarified in Section 2.2 of the initiating event notebook, NB-PSA-IE sequence models with plant identifies how the final support system initiators were

[10]. The final FMEA results were added as Attachment 12 to the initiating damage states; (f) [when identified, how they are grouped, or how the event events notebook.

sequences are modeled tree branches were defined.

Finding Resolved.

using a single top event fault Since these event trees appear to use the same tree] the manner in which the requirements for branches as other event tree, their "grouping" needs accident sequence analysis to be discussed, including the appropriateness of have been satisfied.

using the same event tree nodes for the event trees.

Without a discussion of the event trees and nodes associated with the support system initiators, there is no documentation that the key safety functions or success criteria defined is appropriate and adequate for them.

Although the Controlled Manual Shutdown is listed in the table in Attachment 3, there is no mention of it in the discussion in Section 3. A discussion of the event needs to be included in Section 3 similar to how the other transients are described. Also need to include the actual FMEA in the documentation or provide a valid reference for it.

DOCUMENT the processes While it is obvious that the Flag files exist, the A summary discussion of the event tree rules (Flag) file was added to used to develop accident development and review of the Flag file used in Section 3.0 of the event trees and success criteria notebook NB-PSA-sequences and treat SAPHIRE is not included in the accident sequence ETSC [11]. In addition, Table 3.0-1 was added which lists all c>f the Page 15 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) dependencies in accnient sequences, including the inputs, methods, and results. For example, this documentation typically includes: (a) the linkage between the modeled initiating event in the Initiating Event Analysis section and the accident sequence model; (b) the success criteria established for each modeled initiating event including the bases for the criteria (i.e., the system capacities required to mitigate the accident and the necessary components required to achieve these capacities); (c) a description of the accident progression for each sequence or group of similar sequences (i.e.,

descriptions of the sequence timing, applicable procedural guidance, expected environmental or phenomenological impacts, dependencies between systems and operator actions, end states, and other pertinent information required to fully establish the sequence of events); (d) the operator actions reflected in the event trees, and the sequence-specific timing and dependencies that are traceable to the HRA for these actions; (e) the interface of the accident sequence models with plant analysis documentation. The Flag file does appear to be documented in the Quantification report. However, because this file governs how the accident sequences are quantified, it is important to ensure the accident sequences (especially the support system initiators) are handled correctly in the SAPH IRE model, that the model is modified correctly for applications, and is important for long term maintenance and update of the model. To support this, documentation of the Flag file is an important part of the accident sequence documentation.

It is recommended that Palisades provide at least a brief discussion of the Flag and provide a link to the documentation as it exists in the quantification report.

initiating event logical variables set to true in the rides file for a given initiating transient event. The summary references the detailed discussion for developing event tree rules which is documented in Section 3.0 of the quantification notebook, NB-PSA-QU [6].

Suggestion Resolved.

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition (Supporting Suggestion Category II Text Requirement)

Page 16 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text dependencies in accident analysis documentation. The Flag file does appear to initiating event logical variables set to true in the rules file for a given sequences, including the be documented in the Quantification report. However, initiating transient event. The summary references the detailed discussion inputs, methods, and because this file govems how the accident sequences for developing event tree rules which is documented in Section 3.0 of the results. For example, this are quantified, it is important to ensure the accident quantification notebook, NB-PSA-QU [6].

documentation typically sequences (especially the support system initiators)

Suggestion Resolved.

includes: (a) the linkage are handled correctly in the SAPHIRE model, that the between the modeled model is modified correctly for applications, and is initiating event in the important for long term maintenance and update of Initiating Event Analysis the model. To support this, documentation of the Flag section and the accident file is an important part of the accident sequence sequence model; (b) the documentation.

success criteria established It is recommended that Palisades provide at least a for each modeled initiating event including the bases for brief discussion of the Flag and provide a link to the the criteria (Le., the system documentation as it exists in the quantification report.

capacities required to mitigate the accident and the necessary components required to achieve these capacities); (c) a description of the accident progression for each sequence or group of similar sequences (Le.,

deSCriptions of the sequence timing, applicable procedural guidance, expected environmental or phenomenological impacts, dependencies between systems and operator actions, end states, and other pertinent information required to fully establish the sequence of events); (d) the operator actions reflected in the event trees, and the sequence-specific timing and dependencies that are traceable to the HRA for these actions; (e) the interface of the accident seguence models with plant Page 16 of 55 I

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) damage states; (f) [when sequences are modeled using a single top event fault tree] the manner in which the requirements for accident sequence analysis have been satisfied.

DA-A2-01 Finding ESTABLISH definitions of Component boundaries defined for some Palisades The Palisades PRA modeling intentionally separates contact pairs, SSC boundaries, failure components are not consistent with the generic data breakers etc. from pump motors. This is the correct method of modeling modes, and success criteria component boundaries for the same component. For plant equipment to ensure that appropriate qualitative insights are in a manner consistent with example, the Palisades data report states that the realized. This practice was demonstrated during conduct of the Industry corresponding basic event generic data for motor-driven pumps includes the IREP initiative in 1980 and 1981. Moreover, this practice was adopted in definitions in Systems pump breaker, while the corresponding Palisades the development of the Palisades logic modeling that commenced in 1982 Analysis (SY-A5, SY-A7, component boundary separates the pump breaker in support of the MSIV SEP issue resolution.

SY-A8, SY-A9 through SY-and pump into two separate events, with separate In addition, an evaluation was performed using an interim model (PSAR3 A14 and SY-B4) for failure failure rates for each. This separation also appears to rates and common cause propagate to the definition of component boundaries Release 2b) to determine the magnitude of the potential conservatism failure parameters, and for common cause failures.

introduced by having separate data and component boundaries for ESTABLISH boundaries of breaker-pump combinations as well as other components supplied with unavailability events in a Component boundaries need to be consistent to avoid electrical power via breakers. To bound the problem described in finding manner consistent with potentially double counting failures.

DA-A2-01, a change set was developed with the failure probability for all breakers in the PRA (125 dc, 125 ac, 480 ac, 2400 ac, and 4160 ac) set to corresponding definitions in Keep the separate basic events in the model, but zero. The release 2b base model core damage frequency with the Systems Analysis (SY-A19).

assign a failure probability of 0 to the breaker and normally applied breaker failure probability was 2.29 E-05/yr. With the assign the total failure rate to the pump itself failure probability of all breakers set to zero, the core damage frequency including updating the generic data with the total reduces 19% to 1.85 E-05/yr. This change is less than afactorof two plant-specific failures (pump and associated breaker different and is essentially the same result.

failures), and calculating the corresponding CCF data based on the total failure rate. This allows sensitivities As this difference in total CDF is considered negligible, and the additional and insights to be obtained using the circuit breakers, resolution and insights gained are far more valuable than the while ensuring the model meets the component demonstrated conservatism, it is deemed unnecessary to redefine boundary requirements of the standard. If differences component boundaries in the PRA. In summary, modeling equipment between component boundaries defined in the subcomponents such as breakers, relays, contact pairs, hand switches Palisades PRA and those in generic databases are etc. ensures a comprehensive qualitative characterization of systems, retained, these differences and their bases should be structures and component reliability.

included in the PRA documentation.

Finding Resolved.

DA-C7-01 Finding BASE number of Palisades used actual plant procedures and All preventive maintenance activities for PRA identified components were surveillance tests on plant experience to count surveillance tests. Planned collected from Palisades current equipment database and reviewed.

surveillance requirements maintenance activities are estimated rather than Referring to sections 5.4 of the Data notebook, NB-PSA-DA [5], Palisades Page 17 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion DA-A2-01 Finding DA-C7-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text damage states; (f) [when sequences are modeled using a single top event fault tree] the manner in which the requirements for accident sequence analysis have been satisfied.

ESTABLISH definitions of Component boundaries defined for some Palisades The Palisades PRA modeling intentionally separates contact pairs, SSC boundaries, failure components are not consistent with the generic data breakers etc. from pump motors. This is the correct method of modeling modes, and success criteria component boundaries for the same component. For plant eqUipment to ensure that appropriate qualitative inSights are in a manner consistent with example, the Palisades data report states that the realized. This practice was demonstrated during conduct of the Industry corresponding basic event generic data for motor-driven pumps includes the IREP initiative in 1980 and 1981. Moreover, this practice was adopted in definitions in Systems pump breaker, while the corresponding Palisades the development of the Palisades logic modeling that commenced in 1982 Analysis (SY-A5, SY-A7, component boundary separates the pump breaker in support of the MSIV SEP issue resolution.

SY -A8, SY -A9 through SY-and pump into two separate events, with separate In addition, an evaluation was performed using an interim model (PSAR3 A 14 and SY -84) for failure failure rates for each. This separation also appears to rates and common cause propagate to the definition of component boundaries Release 2b) to determine the magnitude of the potential conservatism failure parameters, and for common cause failures.

introduced by having separate data and component boundaries for ESTABLISH boundaries of breaker-pump combinations as well as other components supplied with unavailability events in a Component boundaries need to be consistent to avoid electrical power via breakers. To bound the problem described in finding manner consistent with potentially double counting failures.

DA-A2-01, a change set was developed with the failure probability for all corresponding definitions in Keep the separate basic events in the model, but breakers in the PRA (125 dc, 125 ac, 480 ac, 2400 ac, and 4160 ac) set to zero. The release 2b base model core damage frequency with the Systems Analysis (SY -A 19).

assign a failure probability of "0* to the breaker and normally applied breaker failure probability was 2.29 E-05/yr. With the assign the "total a failure rate to the pump itself -

failure probability of all breakers set to zero, the core damage frequency including updating the generic data with the atotal" reduces 19% to 1.85 E-05/yr. This change is less than a factor of two plant-specific failures (pump and associated breaker different and is essentially the same result.

failures), and calculating the corresponding CCF data based on the total failure rate. This allows sensitivities As this difference in total CDF is considered negligible, and the additional and insights to be obtained using the circuit breakers, resolution and insights gained are far more valuable than the while ensuring the model meets the component demonstrated conservatism, it is deemed unnecessary to redefine boundary requirements of the standard. If differences component boundaries in the PRA. In summary, modeling equipment between component boundaries defined in the subcomponents such as breakers, relays, contact pairs, hand switches Palisades PRA and those in generic databases are etc. ensures a comprehensive qualitative characterization of systems, retained, these differences and their bases should be structures and component reliability.

included in the PRA documentation.

Finding Resolved.

BASE number of Palisades used actual plant procedures and All preventive maintenance activities for PRA identified components were surveillance tests on plant experience to count surveillance tests. Planned collected from Palisades' current equipment database and reviewed.

surveillance requirement~ _

,---I!!~ntenance activities are estimated rather than Referring to sections 5.4 of the Data notebook, NB-PSA-DA [5], Palisades Page 17 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) and actual practice. BASE being based on maintenance plans.

pm data is based on active, planned PMs. The scope of this project was number of planned three calendar years of plant operation. PM data was actually counted for maintenance activities on preventive maintenance tasks (PMs) with frequencies of three years or plant maintenance plans Contribution from planned maintenance is based on less. For active PMs with frequencies greater than three years, an and actual practice. BASE previous operating experience and not based on equivalency was defined based on the three year data window. For number of unplanned maintenance plans which might be different from the example, if there was a PM that occurred every 6 years, a PM frequency maintenance acts on actual previous plant experience, of 0.5 was assigned. While this number is estimated, it is based on the plant experience, number of actual active, planned PMs. This approach to modeling was reviewed with a qualified PM Program Engineer and validated that this Review planned maintenance activity plans or review approach represents realistic representation of PM frequency.

existing estimates with Maintenance personnel to Documentation of the review was added to Section 5.4 [5].

determine whether estimates of planned maintenance should be changed.

Finding Resolved.

DA-C16-01 Suggestion Data on recovery from loss The SR is met based on a review of data provided in Palisades has re-evaluated the data analysis and the time dependent of offsite power, loss of of NB-PSA-lE (Initiating Event models for the treatment of LOOP events. The modeling aspects include service water, etc. are rare Notebook). Table 9.1 of NB-PSA-IE lists Industry the time of LOOP recovery, the time of onsite power system recovery, on a plant-specific basis. If LOOP Events (1980-2008). The Attachment 9 text EDG mission time, and the coping time between the time of an SBO event available, for each recovery, and Table 9.1 indicate that the August 14, 2003 North and the time when electric power must be recovered to prevent core COLLECT the associated America LOOP events were included (even though it damage. In addition to analyzing these interactions, the August 2003 recovery time with the did not affect Palisades). However, given that this northeast blackout event is evaluated in the data analysis as well.

recovery time being the event was very long for many plants (and very long The approach followed was to consider the probability that re-occurrence period from identification of recoveries significantly affects the LOOP recovery of a regional blackout at Palisades would cause a LOOP as a modeling the system or function distribution), additional discussion of its treatment is uncertainty in the uncertainty analysis using a two stage Bayes failure until the system or appropriate, methodology. The occurrence and non-occurrence of LOOP during the function is returned to In addition, given that other long-term LOOP events NE Blackout at each plant in the industry data was considered based on service.

were screened as not applicable to Palisades, it is whether a LOOP actually occurred or not at each site in the first stage of suggested that a sensitivity analysis address the the Bayes process. In the second stage, two models were effect of the screening process.

probabilistically combined each having a defined probability of being true.

One model assumed the north east Blackout as a Palisades LOOP and Loss of off-site power is an important risk contributor the other counted it as a non-event. Sensitivity studies were performed to and the effect of the screening of longer LOOP events show the effect of altering the probability of LOOP from the base case of in the LOOP recovery analysis have a significant 25% up to 50%. Further details of this analysis are described in Section impact on the risk.

5.7 of NB-PSA-IE [10].

Document treatment of August 14, 2003 North Suggestion Resolved America LOOP event and perform a sensitivity study.

DA-Di-Ol Finding CALCULATE realistic Bayesian updates of all plant specific calculations Palisades parameter estimates are calculated based on Bayesian analysis parameter estimates for used the industry average distributions. For Category employed with a combination of plant specific and generic industry significant basic events II, it is necessary to update significant basic events sources. Generic industry sources include NUREG/CR-6928, Page 18 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion DA-C16-01 Suggestion DA-D1-01 Finding

~------

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text and actual practice. BASE being based on maintenance plans.

pm data is based on active, planned PMs. The scope of this project was number of planned three calendar years of plant operation. PM data was actually counted for maintenance activities on preventive maintenance tasks (PMs) with frequencies of three years or plant maintenance plans Contribution from planned maintenance is based on less. For active PMs with frequencies greater than three years, "an and actual practice. BASE previous operating experience and not based on equivalency was defined based on the three year data window". For number of unplanned maintenance plans which might be different from the example, if there was a PM that occurred every 6 years, a PM frequency maintenance acts on actual previous plant experience.

of 0.5 was assigned." While this number is estimated, it is based on the plant experience.

number of actual active, planned PMs. This approach to modeling was reviewed with a qualified PM Program Engineer and validated that this Review planned maintenance activity plans or review approach represents realistic representation of PM frequency.

existing estimates with Maintenance personnel to Documentation of the review was added to Section 5.4 [5].

determine whether estimates of planned maintenance should be changed.

Finding Resolved.

Data on recovery from loss The SR is met based on a review of data provided in Palisades has re-evaluated the data analysis and the time dependent of offsite power, loss of of NB-PSA-IE (Initiating Event models for the treatment of LOOP events. The modeling aspects include service water, etc. are rare Notebook). Table 9.1 of NB-PSA-IE lists Industry the time of LOOP recovery, the time of onsite power system recovery, on a plant-specific basis. If LOOP Events (1980-2008). The Attachment 9 text EDG mission time, and the coping time between the time of an SBO event available, for each recovery, and Table 9.1 indicate that the August 14, 2003 North and the time when electric power must be recovered to prevent core COLLECT the associated America LOOP events were included (even though it damage. In addition to analyzing these interactions, the August 2003 recovery time with the did not affect Palisades). However, given that this northeast blackout event is evaluated in the data analysis as well.

recovery time being the event was very long for many plants (and very long The approach followed was to consider the probability that re-occurrence period from identification of recoveries significantly affects the LOOP recovery the system or function distribution), additional discussion of its treatment is of a regional blackout at Palisades would cause a LOOP as a modeling failure until the system or appropriate.

uncertainty in the uncertainty analysis using a two stage Bayes' function is retumed to methodology. The occurrence and non-occurrence of LOOP during the service.

In addition, given that other long-term LOOP events NE Blackout at each plant in the industry data was considered based on were screened as not applicable to Palisades, it is whether a LOOP actually occurred or not at each site in the first stage of suggested that a sensitivity analysis address the the Bayes' process. In the second stage, two models were effect of the screening process.

probabilistically combined each having a defined probability of being true.

Loss of off-site power is an important risk contributor One model assumed the north east Blackout as a Palisades LOOP and the other counted it as a non-event. Sensitivity studies were performed to and the effect of the screening of longer LOOP events show the effect of altering the probability of LOOP from the base case of in the LOOP recovery analysis have a significant 25% up to 50010. Further details of this analysis are described in Section impact on the risk.

5.7 of NB-PSA-IE [10].

Document treatment of August 14, 2003 North Suggestion Resolved America LOOP event and perform a sensitivity study.

CALCULATE realistic Bayesian updates of all plant specific calculations Palisades parameter estimates are calculated based on Bayesian analysis parameter estimates for used the industry average distributions. For Category employed with a combination of plant specific and generic industry significant basic events II, it is necessary to update significant basic events sources. Generic industry sources include NUREG/CR-6928, I

I I

I I

Page 18 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) based on relevant generic using a non-informative prior or a prior that represents NUREG-1715 Volume 4, EPRI TR-01 6780 Rev. 6, NUCLARR, and plant-specific evidence the variability in industry data.

NUREG/CR-4639, ASEP, and NUREG/CR-4550, as documented in unless it is justified that 0 of NB-PSA-DA [5]. Plant specific data sources of failure there are adequate plant-For significant components, the use of the industry data included a review of some 10,000 plant work orders and review of specific data to characterize average prior may have distribution spreads that can documented maintenance rule failures as listed in Attachment 3 of [5].

the parameter value and its overwhelm plant experience data when doing a Prior distributions were selected to represent variability in the industry data uncertainty. When it is Bayesian update. Use of the a constrained non-when generic sources were applied. The Bayesian update process was necessary to combine informative prior or a prior reflecting plant to plant performed using the BART code which graphically illustrates the prior and evidence from generic and variability would allow plant operating experience to have a larger impact on the resulting posterior mean.

posterior distributions on the same plot. During this process, there were plant-specific data, USE a no instances during the update of important basic events where it was Bayes update process or Review the significant basic events and evaluate the observed the generic industry data overwhelmed the plant specific data equivalent statistical process plant specific updates based on a constrained non-resulting in a posterior that had very little or no change relative to the prior.

that assigns appropriate informative prior or a prior based plant variability.

Therefore, the use of generic industry distributions in lieu of a weight to the statistical non-informed prior for significant basic events was appropriate.

significance of the generic Finding Resolved.

and plant-specific evidence and provides an appropriate characterization of uncertainty. CHOOSE prior distributions as either noninformative, or representative of variability in industry data.

DA-D4-01 Suggestion When the Bayesian Where generic data was Bayes-updated with plant-All Bayesian update results were reviewed. In cases where there were approach is used to derive a specific data, self-checks should be performed and no plant failures, demand results for means below 1 E-06 and run time distribution and mean value documented to ensure that the posterior distribution rates below 5E-06 were reviewed to ensure they were not unrealistically of a parameter, CHECK that was reasonable.

low. In all cases, changes in the mean were negligible (i.e., less than a the posterior distribution is factor of 2). In cases where there were no plant failures and the failure Based updated data should be confirmed appropriate, rates were above 1 E-06 for demands and 5E-06 for run times, the results reasonable given the relative weight of evidence It is suggested that the data notebook include a were reviewed to confirm the impact from the Bayesian update was provided by the prior and the discussion of how the requirements of this SR DA-D4 minimal (i.e., less than a factor of 3). Attachment 11 [6] provides a plant-specific data.

are met.

comparison of the posterior mean next to the prior.

Examples of tests to ensure In addition, a comparison was made between the data used in the that the updating is accomplished correctly and previous analysis [7] to that used in this update. Failure codes in which there was a measurable difference in the Bayesian updated plant-specific that the generic parameter data (e.g., factor greater than 5) were reviewed in detail. A spreadsheet estimates are consistent with the plant-specific analysis for each Bayesian update was performed using the BART code application include the which provides a visual comparison of the prior and posterior distributions.

following: (a) confirmation The Bayesian update process and reviews performed are described in Section 8.1 of NB-PSA-DA [5].

that_the_Bayesian_updating Page 19 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion DA-D4-01 Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text based on relevant generic using a non-informative prior or a prior that represents NUREG-1715 Volume 4, EPRI TR-016780 Rev. 6, NUCLARR, and plant-specific evidence the variability in industry data.

NUREG/CR-4639, ASEP, and NUREG/CR-4550, as documented in unless it is justified that For significant components, the use of the industry 0 of NB-PSA-DA [5]. Plant specific data sources of failure there are adequate plant-average prior may have distribution spreads that can data included a review of some 10,000 plant work orders and review of specific data to characterize overwhelm plant experience data when doing a documented maintenance rule failures as listed in Attachment 3 of [5].

the parameter value and its Bayesian update. Use of the a constrained non-Prior distributions were selected to represent variability in the industry data uncertainty. When it is informative prior or a prior reflecting plant to plant when generic sources were applied. The Bayesian update process was necessary to combine variability would allow plant operating experience to performed using the BART code which graphically illustrates the prior and evidence from generic and posterior distributions on the same plot. During this process, there were plant-specific data, USE a have a larger impact on the resulting posterior mean.

no instances during the update of important basic events where it was Bayes update process or Review the significant basic events and evaluate the observed the generic industry data overwhelmed the plant specific data equivalent statistical process plant specific updates based on a constrained non-resulting in a posterior that had very little or no change relative to the prior.

that assigns appropriate informative prior or a prior based plant variability.

Therefore, the use of generic industry distributions in lieu of a weight to the statistical non-informed prior for significant basic events was appropriate.

significance of the generic Finding Resolved.

and plant-specific evidence and provides an appropriate characterization of uncertainty. CHOOSE prior distributions as either noninformative, or representative of variability in industry data.

When the Bayesian Where generic data was Bayes-updated with plant-All Bayesian update results were revieWed. In cases where there were approach is used to derive a specific data, self-checks should be performed and no plant failures, demand results for means below 1 E-06 and run time distribution and mean value documented to ensure that the posterior distribution rates below 5E-06 were reviewed to ensure they were not unrealistically of a parameter, CHECK that was reasonable.

low. In all cases, changes in the mean were negligible (Le., less than a the posterior distribution is Based updated data should be confirmed appropriate.

factor of 2). In cases where there were no plant failures and the failure reasonable given the rates were above 1 E-06 for demands and 5E-06 for run times, the results relative weight of evidence It is suggested that the data notebook include a were reviewed to confirm the impact from the Bayesian update was provided by the prior and the discussion of how the requirements of this SR DA-D4 minimal (Le., less than a factor of 3). Attachment 11 [6] provides a plant-specific data.

are met.

comparison of the posterior mean next to the prior.

Examples of tests to ensure In addition, a comparison was made between the data used in the that the updating is accomplished correctly and previous analysis [7] to that used in this update. Failure codes in which that the generic parameter there was a measurable difference in the Bayesian updated plant-specific estimates are consistent data (e.g., factor greater than 5) were reviewed in detail. A spreadsheet with the plant-specific analysis for each Bayesian update was performed using the BART code application include the which provides a visual comparison of the prior and posterior distributions.

following: (a) confirmation The Bayesian update process and reviews performed are described in that the Bayesian updating Section 8.1 of NB-PSA-DA [5].

Page 19 of 55 I

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition (Supporting Suggestion Category II Text Requirement) does not produce a posterior Suggestion resolved.

distribution with a single bin histogram (b) examination of the cause of any unusual (e.g., multimodal) posterior distribution shapes (c) examination of inconsistencies between the prior distribution and the plant-specific evidence to confirm that they are appropriate (d) confirmation that the Bayesian updating algorithm provides meaningful results over the range of values being considered (e) confirmation of the reasonableness of the posterior distribution mean value DA-D8-01 Finding If modifications to plant Plant-specific failure data collected in the collection Section 4.3, Plant Modifications, was added to the Palisades PSA Data design or operating practice data window must be poolable and applicable to the Notebook, NB-PSA-DA [5]. This section of the notebook documents that a lead to a condition where current plant.

review of plant modifications during the data window was performed. A past data are no longer complete list of modifications performed during this time was added to representative of current Only applicable plant-specific data can be applied to of the document.

the failure events.

performance, LIMIT the use of old data: (a)

If the In order to ensure that plant-specific data collected in Finding Resolved.

modification involves new the collection data window is poolable and applicable equipment or a practice to the current plant, plant modifications (both where generic parameter hardware and procedural) implemented during this estimates are available, time period should be reviewed for potential impact USE the generic parameter for on this failure data. This review should be estimates updated with documented and the use of plant-specific data should plant-specific data as it be limited, as appropriate.

becomes available for significant basic events; or (b) If the modification is unique to the extent that generic parameter estimates are not available and only limited_experience_is Page 20 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion DA-DS-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text does not produce a posterior Suggestion resolved.

distribution with a single bin histogram (b) examination of the cause of any unusual (e.g., multimodal) posterior distribution shapes (c) examination of inconsistencies between the prior distribution and the plant-specific evidence to confirm that they are appropriate (d) confirmation that the Bayesian updating algorithm provides meaningful results over the range of values being considered (e) confirmation of the reasonableness of the posterior distribution mean value If modifications to plant Plant-specific failure data collected in the collection Section 4.3, "Plant Modifications", was added to the Palisades PSA Data design or operating practice data window must be poolable and applicable to the Notebook, NB-PSA-DA [5]. This section of the notebook documents that a lead to a condition where current plant.

review of plant modifications during the data window was performed. A past data are no longer Only applicable plant-specific data can be applied to complete list of modifications performed during this time was added to representative of current the failure events. of the document.

performance, LIMIT the use Finding Resolved.

of old data: (a) If the In order to ensure that plant-specific data collected in modification involves new the collection data window is poolable and applicable equipment or a practice to the current plant, plant modifications (both where generic parameter hardware and procedural) implemented during this estimates are available, time period should be reviewed for potential impact USE the generic parameter for on this failure data. This review should be estimates updated with documented and the use of plant-specific data should plant-specific data as it be limited, as appropriate.

becomes available for significant basic events; or (b) If the modification is unique to the extent that generic parameter estimates are not available and only limited experience is Page 20 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

Finding or ASME Reg. Guide 1.200 (Supporting Suggestion Category II Text Finding Description (summary discussion)

Disposition Requirement) available following the change, then ANALYZE the impact of the change and assess the hypothetical effect on the historical data to determine to what extent the data can be used.

HR-Al -01 Finding For equipment modeled in Identification of Pre-Initiator HFEs-No pre-initiator The pre-initiator process was revised to include a process of assessing the PRA, IDENTIFY, HFAs are included for the AFW pump train each system. The initial step of the HFE identification process was to through a review of restoration, common AFW suction from the CST, identify the plant systems to be considered in the review. The Palisades procedures and practices, EDG restoration, High Pressure Safety Injection pre-initiator methodology [8] indicates that the review should include all those test and maintenance (H PSI) pump train restoration, Low Pressure Safety systems modeled in the PRA, which are listed in the Palisades System activities that require Injection (LPSI) pump train restoration, etc. No Notebooks. Once the initial systems list was assembled, the system realignment of equipment documentation was provided on the decision making descriptions and simplified P&lDs were examined to identify and define outside its normal process for excluding restoration errors for standby the Train/Function/Channel (TFC) for the system. Those TFCs not operational or standby components such as these.

susceptible to Type A (pre-initiator) events were screened from further status.

review (this process is documented in Reference [8]). For each of the Restoration of pump train for standby systems can be unscreened TFCs identified, a scoping event was added to the PRA a contributor to risk, model. The scoping values were then used to determine the risk Review each system for possible pre-accident significance of each event and evaluate which events should remain in the restoration errors and if such events are not included model.

in the model, provided a basis for exclusion. The Findin Resolved process identified for screening pre-initiator human g

failure events should be sufficient to identify most pre accident HRAs.

HR-A2-Ol Finding IDENTIFY, through a review Miscalibration events for the containment pressure The Pre-Initiator process was revised to include a process of assessing of procedures and practices, instruments are missing without a detailed screening.

each system. This system level review included potential miscalibration those calibration activities This is similar to the F&O for the restoration events events, including those for the HPSI, LPSI, and containment spray system that if performed incorrectly except for the calibration events. The miscalibration described in this finding. The initial step of the HFE identification process can have an adverse impact events appear to be more complete than the was to identify the plant systems to be considered in the review. The on the automatic initiation of restoration events but additional work is necessary to Palisades pre-initiator methodology indicates that the review should standby safety equipment.

identify the potential miscalibration events, include all systems modeled in the PRA, which are listed in the Palisades System Notebooks. Once the initial systems list was assembled the Miscalibration of containment pressure signals would system descriptions and simplified P&IDs were examined to identify and impact auto start of HPSI,LPSI, and Containment define the Train/Function/Channel (TFC) for the system. Those TFCs not Spray System (CSS). It might also impact auto start of susceptible to Type A (pre-initiator) events were screened from further containment unit coolers and CIS signals.

review. For each of the unscreened TFCs identified, a scoping event was added to the PRA model. The scoping values were then used to Page 21 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion HR-A1-01 Finding HR-A2-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text available follOwing the change, then ANAL YZE the impact of the change and assess the hypothetical effect on the historical data to determine to what extent the data can be used.

For equipment modeled in Identification of Pre-Initiator HFEs-No pre-initiator The pre-initiator process was revised to include a process of assessing the PRA, IDENTIFY, HFAs are included for the AFW pump train each system. The initial step of the HFE identification process was to through a review of restoration, common AFW suction from the CST, identify the plant systems to be considered in the review. The Palisades procedures and practices, EDG restoration, High Pressure Safety Injection pre-initiator methodology [8] indicates that the review should include all those test and maintenance (HPSI) pump train restoration, Low Pressure Safety systems modeled in the PRA, which are listed in the Palisades System activities that require Injection (LPSI) pump train restoration, etc. No Notebooks. Once the initial systems list was assembled, the system realignment of equipment documentation was provided on the decision making descriptions and simplified P&IDs were examined to identify and define outside its normal process for excluding restoration errors for standby the Train/Function/Channel (TFC) for the system. Those TFCs not operational or standby components such as these.

susceptible to Type A (pre-initiator) events were screened from further status.

Restoration of pump train for standby systems can be review (this process is documented in Reference [8]). For each of the a contributor to risk.

unscreened TFCs identified, a scoping event was added to the PRA model. The scoping values were then used to determine the risk Review each system for possible pre-accident significance of each event and evaluate which events should remain in the restoration errors and if such events are not included model.

in the model, provided a basis for exclusion. The Finding Resolved.

process identified for screening pre-initiator human failure events should be sufficient to identify most pre-accident HRAs.

IDENTIFY, through a review Miscalibration events for the containment pressure The Pre-Initiator process was revised to include a process of assessing of procedures and practices, instruments are missing without a detailed screening.

each system. This system level review included potential miscalibration those calibration activities This is similar to the F&O for the restoration events events, including those for the HPSI, LPSI, and containment spray system that if performed incorrectly except for the calibration events. The miscalibration described in this finding. The initial step of the HFE identification process can have an adverse impact events appear to be more complete than the was to identify the plant systems to be considered in the review. The on the automatic initiation of restoration events but additional work is necessary to Palisades pre-initiator methodology indicates that the review should standby safety equipment.

identify the potential miscalibration events.

include all systems modeled in the PRA, which are listed in the Palisades Miscalibration of containment pressure Signals would System Notebooks. Once the initial systems list was assembled, the system descriptions and simplified P&IDs were examined to identify and impact auto start of HPSI, LPSI, and Containment define the Train/Function/Channel (TFC) for the system. Those TFCs not Spray System (CSS). It might also impact auto start of susceptible to Type A (pre-initiator) events were screened from further containment unit coolers and CIS signals.

review. For each of the unscreened TFCs identified, a scoping event was added to the PRA model. The scoping values were then used to Page 21 of 55

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Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions irting 1.200 Finding Description (summary discussion)

Disposition Requirement)

Review each system for possible pre-accident determine the risk significance of each event and evaluate which events restoration errors and if such events are not included should remain in the model. For each of the HFEs that were added to the in the model, provided a basis for exclusion. The model a more detailed assessment of the system design and the process identified for screening pre-accident human procedures governing the maintenance, surveillance and testing of the actions should be sufficient to identify most pre-associated components was performed in order to support a detailed accident HRAs.

assessment of the HEP. The detailed methodology is described in the updated HRA Notebook Volume 2 NB-PSA-HR Volume 2, Palisades Pre-Initiator Human Error Evaluation [8].

Finding Resolved.

HR-C1-01 Suggestion For each unscreened Many of the pre-initiator human failure events The Pre-Initiator process was revised to include a process of assessing activity, DEFINE a human identified in Tables E.2-1A and E.2-1B do not match each system. The initial step of the HFE identification process was to failure event (HFE) that the basic event name in the Palisades fault tree. For identify the plant systems to be considered in the review. The Palisades represents the impact of the these events, it appears that the system designator pre-initiator methodology indicates that the review should include all human failure at the has been expanded from one character to three systems modeled in the PRA, which are listed in the Palisades System appropriate level, i.e.,

characters in the BE name.

Notebooks. Once the initial systems list was assembled, the system function system train, or descriptions and simplified P&IDs were examined to identify and define component affected.)

Inconsistencies between the documentation and the the Train/Function/Channel (TFC) for the system. Those TFCs not model make reviews difficult and might lead to susceptible to Type A (pre-initiator) events were screened from further additional questions on mode a equacy.

review (this process is documented in Table 2.2-1 of NB-PSA-HR Volume Update the HRA evaluations and the HRA document II [8]). For each of the unscreened TFCs identified, a scoping event was to match the BE5 listed in the fault tree.

added to the PRA model. The scoping values were then used to determine the risk significance of each event and evaluate which events should remain in the model.

The basis for any exclusion of pre-initiator for a system is documented in the updated HRA Notebook. A specific check to confirm that basic events representing the pre-initiators was performed to confirm events in the PRA model agree with the development discussed in the HRA notebook (NB PSA-HR Volume II [8]).

Suggestion Resolved HR-C2-01 Finding INCLUDE those modes of Pre-initiator human failure events were included in the The pre-initiator methodology was revised and each system re-evaluated unavailability that, following fault tree at the appropriate level for the pre-initiator for the possibility that pre-initiator events could occur at the completion of each HFE identified. However, based on the missing HFE5 train/channel/function level of each system. The revised methodology and unscreened activity, result identified in F&Os against HR-Al and HR-A2, and no new pre-initiator HEPs are discussed in the HRA Notebook Volume II [8].

from failure to restore (a) evidence of a review of plant specific mispositioning Those pre-initiators specifically identified during the review and several equipment to the desired or miscalibration events, credit cannot be given for others were assessed using the revised methodology and, as necessary, standby or operational collection of plant-specific or generic operating were added to the model. A review of plant history was conducted for status (b) initiation signal or experience, plant specific operating experience. The result of the review was that set point for equipment start-while there were instances noted of conditions that would be considered No_review_of_plant_misposition_or_miscalibration_and Page 22 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion HR-C1-01 Suggestion HR-C2-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text Review each system for possible pre-accident determine the risk significance of each event and evaluate which events restoration errors and if such events are not included should remain in the model. For each of the HFEs that were added to the in the model, provided a basis for exclusion. The model a more detailed assessment of the system design and the process identified for screening pre-accident human procedures goveming the maintenance, surveillance and testing of the actions should be sufficient to identify most pre-associated components was performed in order to support a detailed accident HRAs.

assessment of the HEP. The detailed methodology is described in the updated HRA Notebook Volume 2 NB-PSA-HR Volume 2, "Palisades Pre-Initiator Human Error Evaluation" (8).

Finding Resolved.

For each unscreened Many of the pre-initiator human failure events The Pre-Initiator process was revised to include a process of assessing activity, DEFINE a human identified in Tables E.2-1 A and E.2-1 B do not match each system. The initial step of the HFE identification process was to failure event (HFE) that the basic event name in the Palisades fault tree. For identify the plant systems to be considered in the review. The Palisades represents the impact of the these events, it appears that the system deSignator pre-initiator methodology indicates that the review should include all human failure at the has been expanded from one character to three systems modeled in the PRA, which are listed in the Palisades System appropriate level, i.e.,

characters in the BE name.

Notebooks. Once the initial systems list was assembled, the system function, system, train, or Inconsistencies between the documentation and the descriptions and simplified P&IDs were examined to identify and define component affected.)

model make reviews difficult and might lead to the Train/FunctioniChannel (TFC) for the system. Those TFCs not susceptible to Type A (pre-initiator) events were screened from further additional questions on model adequacy.

review (this process is documented in Table 2.2-1 of NB-PSA-HR Volume Update the HRA evaluations and the HRA document II [8)). For each of the unscreened TFCs identified, a scoping event was to match the BEs listed in the fault tree.

added to the PRA model. The scoping values were then used to determine the risk significance of each event and evaluate which events should remain in the model.

The basis for any exclusion of pre-initiator for a system is documented in the updated HRA Notebook. A specific check to confirm that basic events representing the pre-initiators was performed to confirm events in the PRA model agree with the development discussed in the HRA notebook (NB-PSA-HR Volume II (8)).

Suggestion Resolved INCLUDE those modes of Pre-initiator human failure events were included in the The pre-initiator methodology was revised and each system re-evaluated unavailability that, following fault tree at the appropriate level for the pre-initiator for the possibility that pre-initiator events could occur at the completion of each HFE identified. However, based on the missing HFEs trainlchanneVfunction level of each system. The revised methodology and unscreened activity, result identified in F&Os against HR-A 1 and HR-A2, and no new pre-initiator HEPs are discussed in the HRA Notebook Volume II [8).

from failure to restore (a) evidence of a review of plant specific mispositioning Those pre-initiators specifically identified during the review and several equipment to the desired or miscalibration events, credit cannot be given for others were assessed using the revised methodology and, as necessary, standby or operational collection of plant-specifiC or generic operating were added to the model. A review of plant history was conducted for status (b) initiation signal or experience.

plant specific operating experience. The result of the review was that set point for equipment start-No reviELw_ of plant misp~ition or miscalibration ancj while there were instances noted of conditions that would be considered Page 22 of 55

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Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) up or realignment (c) missing events generally included for standby pre-initiators, the examples noted were either already covered by a automatic realignment or components and instrumentation as discussed in HR-pre-initiator event identified during the implementation of the revised power ADD failure modes Al and HR-A2.

methodology or were related to equipment not credited in the PRA. The identified during the plant operating experience review is documented in HRA notebook Perform a systematic review of HFEs. Consider a volume 1 [25].

collection of plant-specific or Condition Report review of mispositioned or applicable generic operating miscalibrated events to determine if any trends Finding Resolved.

experience that leave associated with the pre-accident events could impact equipment unavailable for the HRA values.

response in accident sequences.

HR-D4-0l Suggestion When taking into account Recovery factors are credited in the detailed The HFE summary table, Table 4-1 of NB-PSA-HR Vol. 2 [8], was revised self-recovery or recovery evaluations of pre-initiator human failure events to provide a listing of the recovery factor ASEP cases applied to from other crew members in (HFEs). However, the write-up is not clear which of pre-initiators HFEs for which detailed analysis was performed.

estimating HEPs for specific the ASEP cases that the evaluation represents. The HFEs, USE pre-initiator HRA Calculator shows the associated ASEP case in Suggestion Resolved recovery factors in a manner the detailed evaluation. However, the information is consistent with selected not presented in the document which makes review methodology. If recovery of somewhat more difficult.

pre-initiator errors is credited This is a documentation issue only.

(a) ESTABLISH the maximum credit that can be Update the HRA calculator write-up to be clearer as to given for multiple recovery the ASEP case used for the detailed evaluations.

opportunities (b) USE the following information to assess the potential for recovery of pre-initiator: (1) post-maintenance or post-calibration tests required and performed by procedure (2) independent verification, using a written check-off list, that verifies component status following maintenance/testing (3) a separate check of component status made at a later time, using a written check-off list, by the original performer (4) work shift or daily checks of component status, using a written Page 23 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion HR-D4-01 Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text up or realignment (c) missing events generally included for standby pre-initiators, the examples noted were either already covered by a automatic realignment or components and instrumentation as discussed in HR-pre-initiator event identified during the implementation of the revised power ADD failure modes A1 and HR-A2.

methodology or were related to equipment not credited in the PRA. The identified during the Perform a systematic review of HFEs. Consider a plant operating experience review is documented in HRA notebook collection of plant-specific or Condition Report review of mispositioned or volume 1 [251.

applicable generic operating miscalibrated events to determine if any trends Finding Resolved.

experience that leave equipment unavailable for associated with the pre-accident events could impact response in accident the HRA values.

sequences.

When taking into account Recovery factors are credited in the detailed The HFE summary table, Table 4-1 of NB-PSA-HR Vol. 2 [81, was revised self-recovery or recovery evaluations of pre-initiator human failure events to provide a listing of the recovery factor ASEP cases applied to from other crew members in (HFEs). However, the write-up is not clear which of pre-initiators HFEs for which detailed analysis was performed.

estimating HEPs for specific the ASEP cases that the evaluation represents. The Suggestion Resolved HFEs, USE pre-initiator HRA Calculator shows the associated ASEP case in recovery factors in a manner the detailed evaluation. However, the information is consistent with selected not presented in the document which makes review methodology. If recovery of somewhat more difficult.

pre-initiator errors is credited This is a documentation issue only.

(a) ESTABLISH the maximum credit that can be Update the HRA calculator write-up to be clearer as to given for multiple recovery the ASEP case used for the detailed evaluations.

opportunities (b) USE the following information to assess the potential for recovery of pre-initiator: (1) post-maintenance or post-calibration tests required and performed by procedure (2) independent verification, using a written check-off list, that verifies component status following maintenance/testing (3) a separate check of component status made at a later time, using a written check-off list, by the original performer (4) work shift or daily checks of component status, using a written

~~---

Page 23 of 55 I

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Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions (Supporting 1.200 Finding Description (summary discussion)

Disposition Requirement) check-off list.

HR-E1-01 Suggestion When identifying the key There does not appear to be a systematic review of Procedures have routinely been reviewed by system analysts and HRA human response actions procedures and other relevant guidance in the analysts throughout the historical development of the PRA model.

REVIEW: (a) the plant-identification of post-accident human errors. Cutset Palisades previously and currently has on staff an individual with prior specific emergency reviews and expert panel review have likely identified plant specific experience as an SRO. This individual is responsible for operating procedures, and any significant post-accident HRAs that were not developing input to new HFEs and updating existing HFE5 via the other relevant procedures previously modeled. However, the system notebook development of a Post-Initiator Operator Action Questionnaire (PIOAQ).

(e.g., AOPs, annunciator review of the operator actions does not include ties to The purpose of the PIOAQ is to document from an operators perspective response procedures) in the the procedures that indicates an HRA review per SY-of the expected response to identified events and specified event context of the accident Al 7.

scenarios. The individual documents on the form the expected scenarios (b) system progression of the operator response from event initiation and the operation such that an procedure(s) the operator would be following in the event response.

understanding of how the The need to include these HFEs is driven by review of the quantified system(s) functions and the human interfaces with the model results following completion of changes to or updates of the model.

s stem is obtained The procedural guidance is documented in both the PIOAQ form and in y

the development of the HEP in the HRA calculator.

The system notebooks include a general discussion of HFEs included in the system model and a detailed listing of each HFE in the system model.

In addition, the reference section of each notebook includes those procedures which identify the system capability and operation. These references include applicable Emergency Operating Procedures (EOP),

System Operating Procedures (SOP), Off-Normal Procedures, Alarm Response Procedures (ARP), etc. which establish the guidance for expected operator action to maintain system operation or removal from service and the expected operator action in response to identified events.

Suggestion Resolved HR-E1-02 Suggestion When identifying the key In Table 2-1 HRA Summary Table, for Type C:

The guidance provided in the HRA notebooks indicates that the penalty of human response actions Procedural Actions During Course of Accident, it a 1.0 human error probability (HEP) for these actions is overly REVIEW: (a) the plant-states that the evaluation of actions with no conservative for actions where the time to complete the action is long and specific emergency procedures can be performed using EPRI TR-100259.

additional resources are available (TSC, OSC. etc.). The suggestion is operating procedures, and NUREG-1335 suggests a value of 1.0 for non-that since the guidance hasnt actually been applied to any current HEP other relevant procedures proceduralized actions; however, this is judged to be values, the guidance itself should be removed to avoid confusion. The (e.g., AOPs, annunciator overly conservative when the time available for action purpose of this wording in the notebook is to provide guidance for a variety response procedures) in the is long and the TSC becomes available.

of circumstances that may or may not be applicable at any given time. As context of the accident the guidance was not implemented at the time of the PEER review does scenarios (b) system However, it doesn t appear that any operator actions not mean it could not be implemented later.

operation such that an Page 24 of 55 F&O#

Finding or (Supporting Suggestion Requirement)

HR-E1-01 Suggestion HR-E1-02 Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Category II Text Finding Description (summary discussion)

Disposition check*off list.

When identifying the key There does not appear to be a systematic review of Procedures have routinely been reviewed by system analysts and HRA human response actions procedures and other relevant guidance in the analysts throughout the historical development of the PRA model.

REVIEW: (a) the plant-identification of post-accident human errors. Cutset Palisades previously and currently has on staff an individual with prior specific emergency reviews and expert panel review have likely identified plant specific experience as an SRO. This individual is responsible for operating procedures, and any significant post-accident HRAs that were not developing input to new HFEs and updating existing HFEs via the other relevant procedures previously modeled. However, the system notebook development of a Post-Initiator Operator Action Questionnaire (PIOAQ).

(e.g., AOPs, annunciator review of the operator actions does not include ties to The purpose of the PIOAQ is to document from an operators' perspective response procedures) in the the procedures that indicates an HRA review per SY-of the expected response to identified events and specified event context of the accident A17.

scenarios. The individual documents on the form the expected scenarios (b) system progression of the operator response from event initiation and the operation such that an procedure(s) the operator would be follOwing in the event response.

understanding of how the The need to include these HFEs is driven by review of the quantified system(s) functions and the human interfaces with the model results following completion of changes to or updates of the model.

system is obtained The procedural guidance is documented in both the PIOAQ form and in the development of the HEP in the HRA calculator.

The system notebooks include a general discussion of HFEs included in the system model and a detailed listing of each HFE in the system model.

In addition, the reference section of each notebook includes those procedures which identify the system capability and operation. These references include applicable Emergency Operating Procedures (EOP),

System Operating Procedures (SOP), Off-Normal Procedures, Alarm Response Procedures (ARP), etc. which establish the guidance for expected operator action to maintain system operation or removal from service and the expected operator action in response to identified events.

Suggestion Resolved When identifying the key In Table 2-1 HRA Summary Table, for Type C:

The guidance provided in the HRA notebooks indicates that the penalty of human response actions Procedural Actions During Course of Accident, it a 1.0 human error probability (HEP) for these actions is overly REVIEW: (a) the plant-states that Uthe evaluation of actions with no conservative for actions where the time to complete the action is long and specific emergency procedures can be performed using EPRI TR-100259.

additional resources are available (TSC, OSC. etc.). The suggestion is operating procedures, and NUREG-1335 suggests a value of 1.0 for non-that since the guidance hasn't actually been applied to any current HEP other relevant procedures proceduralized actions; however, this is judged to be values, the guidance itself should be removed to avoid confusion. The I

(e.g., AOPs, annunciator overly conservative when the time available for action purpose of this wording in the notebook is to provide guidance for a variety response procedures) in the is long and the TSC becomes available.*

of circumstances that mayor may not be applicable at any given time. As context of the accident However, it doesn't appear that any operator actions the guidance was not implemented at the time of the PEER review does scenarios (b) system not mean it could not be implemented later.

operation such that an used this.

Page 24 of 55

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Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition (Supporting Suggestion category II Text Requirement) understanding of how the This is primarily a documentation issue. Statement Suggestion Resolved system(s) functions and the causes contusion if not used.

human interfaces with the system is obtained Although non-proceduralized operator actions dont appear to be credited, it is suggested that this statement be removed to avoid confusion about the credit for non-proceduralized operator actions.

HR-E3-01 Finding TALK THROUGH (i.e.,

HRAs were reviewed by former SRO to ensure and A copy of the Human Error Probability (HEP) Post-Initiator Calculations review in detail) with plant confirm that interpretation of the procedures is (P-IC) and associated Post-Initiator Operator Action Questionnaire operations and training consistent with plant observations and training (P-IOAQ) were provided to current SRO licensed on-shift Operations personnel the procedures procedures. No review by training personnel was Department personnel and Training Department personnel for use in and sequence of events to performed as required by Cat II & Ill.

validating HEP information accuracy.

confirm that interpretation of the procedures is consistent No review by training personnel was performed as HEPs were assigned to the five Operations Department Operating Crews required by Cat II & Ill.

(10 per crew) for review. Their reviews included ensuring indications, with plant observations and procedure selection and use, and activity performance man-power and training procedures.

Document the talk through performed with training personnel to confirm the interpretation of the timing is correct. Training personnel reviews included ensuring procedure selection and use were consistent with current training expectations, and procedures is consistent with plant observations and the training type and frequency are accurate.

training procedures.

Operator comments were reviewed and proposed resolutions forwarded to the comment initiator for further comment or acceptance. Comment initiator acceptance is documented by their initialing the HEP Validation form.

The records of the current operating crews and training personnel are provided in Attachment F of the HRA notebook volume 1 [25].

Finding Resolved.

HR-G6-01 Finding CHECK the consistency of HRA Procedure 5.3.2.12 states: The consistency of A comparison of the human error probabilities (HEPs) developed for each the post-initiator HEP resulting post-initiator Human Error Probabilities human failure event (HFE) in the PLP internal events PRA model shows quantifications. REVIEW the (HEP5) should be checked: (a) REVIEW the Human that the values of the HEPs are internally consistent relative to each other, HFEs and their final HEPs Failure Events (HFE5) and their final HEPs relative to and generally follow a trend of lower HEPs being associated with lower relative to each other to each other to check their reasonableness given the stress levels (which in turn may be associated with more time available to check their reasonableness scenario context, plant history, procedures, take action). Exceptions to the general trend are present, and can be given the scenario context, operational practices, and experience. (b) One explained through detailed examinations of the contributions to the HEP plant history, procedures, approach for checking the consistency of HEP (e.g., number of procedure steps, time available to perform the steps, operational practices, and quantifications is to sort by increasing or decreasing probability of successfully recovering from errors occurring during experience.

HEP values and then performing the comparison. In completion of the procedure, etc.). This review is documented in Section addition, HRA Notebook Section 4.0 states: After the 4.0 of NB-PSA-HR Volume 1 [25].

individual results were obtained, the final HEPs were assessed for appropriateness and consistency within Finding Resolved.

Page 25 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion HR-E3-01 Finding HR-G6-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text understanding of how the This is primarily a documentation issue. Statement Suggestion Resolved system(s) functions and the causes confusion if not used.

human interfaces with the Although non-proceduralized operator actions don't system is obtained appear to be credited, it is suggested that this statement be removed to avoid confusion about the credit for non-proceduralized operator actions.

TALK THROUGH (Le.,

HRAs were reviewed by former SRO to ensure and A copy of the Human Error Probability (HEP) Post-Initiator Calculations review in detail) with plant confirm that interpretation of the procedures is (P-IC) and associated Post-Initiator Operator Action Questionnaire operations and training consistent with plant observations and training (P-IOAQ) were provided to current SRO licensed on-shift Operations personnel the procedures procedures. No review by training personnel was Department personnel and Training Department personnel for use in and sequence of events to performed as required by Cat II & III.

validating HEP information accuracy.

confirm that interpretation of No review by training personnel was performed as HEPs were assigned to the five Operations Department Operating Crews the procedures is consistent with plant observations and required by Cat II & III.

(-10 per crew) for review. Their reviews included ensuring indications, training procedures.

Document the talk through performed with training procedure selection and use, and activity performance man-power and timing is correct. Training personnel reviews included ensuring procedure personnel to confirm the interpretation of the selection and use were consistent with current training expectations, and procedures is consistent with plant observations and the training type and frequency are accurate.

training procedures.

Operator comments were reviewed and proposed resolutions forwarded to the comment initiator for further comment or acceptance. Comment initiator acceptance is documented by their initialing the HEP Validation form.

The records of the current operating crews and training personnel are provided in Attachment F of the HRA notebook volume 1 [25].

Finding Resolved.

CHECK the consistency of HRA Procedure 5.3.2.12 states: "The consistency of A comparison of the human error probabilities (HEPs) developed for each the post-initiator HEP resulting post-initiator Human Error Probabilities human failure event (HFE) in the PLP internal events PRA model shows quantifications. REVIEW the (HEPs) should be checked: (a) REVIEW the Human that the values of the HEPs are internally consistent relative to each other, HFEs and their final HEPs Failure Events (HFEs) and their final HEPs relative to and generally follow a trend of lower HEPs being associated with lower relative to each other to each other to check their reasonableness given the stress levels (which in tum may be associated with more time available to check their reasonableness scenario context, plant history, procedures, take action). Exceptions to the general trend are present, and can be given the scenario context, operational practices, and experience. (b) One explained through detailed examinations of the contributions to the HEP plant history, procedures, approach for checking the consistency of HEP (e.g., number of procedure steps, time available to perform the steps, operational practices, and quantifications is to sort by increasing or decreasing probability of successfully recovering from errors occurring during experience.

HEP values and then performing the comparison.* In completion of the procedure, etc.). This review is documented in Section addition, HRA Notebook Section 4.0 states: "After the 4.0 of NB-PSA-HR Volume 1 [25].

individual results were obtained, the final HEPs were Finding Resolved.

assessed for appropriateness and consistency within Page 25 of 55

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Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition (Supporting Suggestion Category II Text Requirement) the PLP HRA. Human action element such as time frame and complexity of the actions diagnosis and/or execution were considered. When available, HEP results for similar actions at other PWRs were used as further points of reference. However, no documentation of the review was found.

Consistency check is required for Capability Category I, II, and Ill. Document the consistency check that was performed.

HR-G7-01 Finding For multiple human actions Palisades has not completed their HFE Dependency The methodology for evaluating human error dependency was developed in the same accident Evaluation for their updated HRA. This is specifically as described in HRA Notebook NB-PSA-HR Volume 1 [25].

sequence or cut set, noted in Section 5.2 of PLP-HRA.

This approach evaluates the dependency between the multiple operator identified in accordance with Failure to meet explicit requirement of the standard.

actions that occur in the accident sequences of the Palisades PSA. The supporting requirement QU-human reliability analysis of the PSA developed human error probabilities Cl, ASSESS the degree of After the HRA is complete, redo and document the (HEP5) as though they were independent of one another.

It is known that dependence, and calculate dependency evaluation, a number of these operator actions appear in the same accident a joint human error sequences. If dependencies exist between these operator actions, then probability that reflects the the core damage frequency may be higher than quantified in the accident dependence. ACCOUNT for sequence analysis. This analysis will evaluate the post-initiator the influence of success or dependencies among operator actions credited in the Palisades PSA and failure in preceding human determines whether the impact of these dependencies on the overall core actions and system damage frequency is significant. The most risk significant human error performance on the human dependencies will be fully developed into conditional human actions and event under consideration incorporated explicitly in the Palisades PSA fault trees.

including (a) the time required to complete all The general steps used in this analysis are as follows:

actions in relation to the time

1. Run the base model with the post-initiator action failure event available to perform the probabilities set to 1.0.

actions (b) factors that could lead to dependence (e.g.,

2. Identify the multiple human action combinations that appear in the cut common instrumentation, sets.

common procedures,

3. Identify the risk significant combinations assuming complete increased stress, etc.) (c) dependence.

availability of resources (e.g., personnel)

4. Perform a dependency analysis on the risk significant combinations and develop conditional probabilities for dependent actions.

5. Incorporate the dependent combinations in the fault trees of the PSA.

To address the human action dependency issue with respect to CDF, Palisades_developed_a_systematic_approach_that_investigates_a_sufficient Page 26 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion HR-G7-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text the PLP HRA. Human action element such as time frame and complexity of the action's diagnosis and/or execution were considered. When available, HEP results for similar actions at other PWRs were used as further points of reference." However, no documentation of the review was found.

Consistency check is required for Capability Category I, II, and III. Document the consistency check that was performed.

For multiple human actions Palisades has not completed their HFE Dependency The methodology for evaluating human error dependency was developed in the same accident Evaluation for their updated HRA. This is specifically as described in HRA Notebook NB-PSA-HR Volume 1 [25].

sequence or cut set, noted in Section 5.2 of PLP-HRA.

This approach evaluates the dependency between the multiple operator identified in accordance with supporting requirement au-Failure to meet explicit requirement of the standard.

actions that occur in the accident sequences of the Palisades PSA. The C1, ASSESS the degree of After the HRA is complete, redo and document the human reliability analysis of the PSA developed human error probabilities dependence, and calculate dependency evaluation.

(HEPs) as though they were independent of one another. It is known that a joint human error a number of these operator actions appear in the same accident probability that reflects the sequences. If dependencies exist between these operator actions, then dependence. ACCOUNT for the core damage frequency may be higher than quantified in the accident the influence of success or sequence analysis. This analysis will evaluate the post-initiator failure in preceding human dependencies among operator actions credited in the Palisades PSA and actions and system determines whether the impact of these dependencies on the overall core performance on the human damage frequency is Significant. The most risk significant human error event under consideration dependencies will be fully developed into conditional human actions and including (a) the time incorporated explicitly in the Palisades PSA fault trees.

required to complete all The general steps used in this analysis are as follows:

actions in relation to the time

1. Run the base model with the post-initiator action failure event available to perform the actions (b) factors that could probabilities set to 1.0.

lead to dependence (e.g.,

2. Identify the multiple human action combinations that appear in the cut common instrumentation, sets.

common procedures,

3. Identify the risk significant combinations assuming complete increased stress, etc.) (c) availability of resources dependence.

(e.g., personnel)

4. Perform a dependency analysis on the risk significant combinations and develop conditional probabilities for dependent actions.

5. Incorporate the dependent combinations in the fault trees of the PSA.

To address the human action dependency issue with respect to CDF, Palisades developed a systematic approach that investigates a sufficient Page 26 of 55

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Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Requirement)

Suggestion Category II Text number of human actions to merit confidence that the impact of these dependencies have been thoroughly assessed and adequately represented in the PSA models. The approach is iterative and methodical.

The results of this dependency evaluation will be reflected through explicit modeling of dependencies within the fault tree models.

It is expected the resulting CDF will be slightly increased over that developed using independent (zero dependence) event combinations.

A human error dependency analysis was completed for the flooding PRA as documented in EA-PSA-INTFLOOD-13-06 Rev. 0 [43]. This finding is still open for the full power internal events PRA.

Finding Open for full power internal events HEPs.

HR-Hi -01 Suggestion INCLUDE operator recovery ESS-XVOT-SWS-ESS was added as an example of The task described in this suggestion is performed as part of normal HEP actions that can restore the how Palisades applied operator recovery actions that development preparation and documentation. Significant sequences are functions, systems, or can restore components on an as-needed basis to reviewed on every PRA model update and change.

If a sequence is components on an as-provide a more realistic evaluation of significant deemed overly conservative due to potential operator action that is not needed basis to provide a accident sequences. However, the performance of the currently credited, then a new HEP is developed based on plant more realistic evaluation of review of significant sequences to determine if procedures and cues that occur dunng the sequence. The review of significant accident recovery actions are needed was not documented.

significant sequences for purposes of selecting which could be made more sequences.

realistic is not required to be explicitly documented per the standard SR.

The documentation of the performance of the review of significant sequences to determine if recovery Suggestion Resolved actions are needed is required to continue to meet this SR.

Consider adding documentation of this review to the next major PRA model update.

HR-l3-0i Finding DOCUMENT the sources of There are only two assumptions in the entire HRA Table 1.6.1 was added to the Human Reliability Analysis Notebook NB-model uncertainty and notebook. Both are associated with individual HRAs.

PSA-HR Volume 1 [25]. This table documents over 60 assumptions related assumptions (as General assumptions associated with HRA minimum including basis, assumption type, and model uncertainty impact. The identified in QU-El and QU-defaults and methodology requirements are not listed assumptions are categorized into fire related and general HRA methods E2) associated with the as assumptions and are thus not addressed in terms assumptions.

human reliability analysis.

of model uncertainty.

Finding Resolved.

Only two assumptions were listed for all of the HRAs.

This does not appear to be consistent with the remainder of the model in terms of assumptions and sources of model uncertainty.

Review the HRA for additional imbedded assumptions and_use_the_updated_list_for_potential_model Page 27 of 55 F&O#

(Supporting Finding or Requirement)

Suggestion HR-H1-01 Suggestion HR-13-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text number of human actions to merit confidence that the impact of these dependencies have been thoroughly assessed and adequately represented in the PSA models. The approach is iterative and methodical.

The results of this dependency evaluation will be reflected through explicit modeling of dependencies within the fault tree models. It is expected the resulting CDF will be slightly increased over that developed using independent (zero dependence) event combinations.

A human error dependency analysis was completed for the flooding PRA as documented in EA-PSA-INTFLOOD-13-06 Rev. 0 [43]. This finding. is still open for the full power intemal events PRA.

Finding Open for full power intemal events HEPs.

INCLUDE operator recovery a ESS-XVOT -SWS-ESS* was added as an example of The task described in this suggestion is performed as part of normal HEP actions that can restore the how Palisades applied operator recovery actions that development preparation and documentation. Significant sequences are functions, systems, or can restore components on an as-needed basis to reviewed on every PRA model update and change. If a sequence is components on an as-provide a more realistic evaluation of significant deemed overly conservative due to potential operator action that is not needed basis to provide a accident sequences. However, the performance of the currently credited, then a new HEP is developed based on plant more realistic evaluation of review of Significant sequences to determine if procedures and cues that occur during the sequence. The review of Significant accident recovery actions are needed was not documented.

significant sequences for purposes of selecting which could be made more sequences.

The documentation of the performance of the review realistic is not required to be explicitly documented per the standard SA.

of significant sequences to determine if recovery Suggestion Resolved actions are needed is required to continue to meet this SA.

Consider adding documentation of this review to the next major PRA model update.

DOCUMENT the sources of There are only two assumptions in the entire HRA Table 1.6.1 was added to the Human Reliability Analysis Notebook NB-model uncertainty and notebook. Both are associated with individual HRAs.

PSA-HR Volume 1 [25]. This table documents over 60 assumptions related assumptions (as General assumptions associated with HRA minimum including basis, assumption type, and model uncertainty impact. The identified in aU-E1 and au-defaults and methodology requirements are not listed assumptions are categorized into fire related and general HRA methods E2) associated with the as assumptions and are thus not addressed in terms assumptions.

human reliability analysis.

of model uncertainty.

Finding Resolved.

Only two assumptions were listed for all of the HRAs.

This does not appear to be consistent with the remainder of the model in terms of assumptions and sources of model uncertainty.

Review the HRA for additional imbedded assumptions and use the updated list for potential model J

I Page 27 of 55

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Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) uncertainties.

IE-A6-01 Finding When performing the Although it appears that an evaluation of CCFs was A detailed evaluation to address this finding was completed in Attachment systematic evaluation performed since IE_LOY1 0-Y20 and 3 of Reference [9]. In summary, the evaluation states:

required in IE-A5, INCLUDE lE_LO-ALL4PREFAC, etc, were identified; however, initiating events resulting documentation of the systematic evaluation for the A process to ensure all possible common cause combinations and routine and non-routine system alignments is theoretically achievable, but time from multiple failures, if the elimination of other support system CCF events was equipment failures result not provided, consuming and inconsistent with risk-informed approaches as utilized in PRAs. However, a process to ensure that combinations of failure events from a common cause, and Provide documentation of the evaluation of electrical and system configurations that have occurred or could reasonably occur is from routine system equipment CCF initiating events, in place already through the current Palisades approach of considering alignments, plant and generic data, initiating event categorization, and technical specifications. This process addresses reasonable common cause combinations if in fact such combinations are necessary to result in a plant trip.

Finding Resolved.

IE-A6-02 Finding When performing the Event trees for common cause failures (e.g., Loss of A detailed evaluation to address this finding was completed in Attachment systematic evaluation Preferred AC Bus Y20, Y30, and Y40) are included in 3 of Reference [9].

required in IE-A5, INCLUDE the SAPHIRE program, but no documentation In summary, the evaluation states:

initiating events resulting associated with these event trees has been found.

from multiple failures, if the Note, the FMEA discussion provided in the Initiating A process to ensure all possible common cause combinations and routine equipment failures result Event notebook, does not specifically discuss the and non-routine system alignments is theoretically achievable, but time from a common cause, and CCF initiators, nor does it identify the buses as consuming and inconsistent with risk-informed approaches as utilized in from routine system necessarily resulting in a reactor trip. No discussion of PRAs. However, a process to ensure that combinations of failure events alignments, non-routine system alignments has been found.

and system configurations that have occurred or could reasonably occur is The NRCs clarification for this element requires in place already through the current Palisades approach of considering consideration, and documentation of Initiating events plant and genenc data, initiating event categonzation, and technical resulting from common cause or from both routine specifications. This process addresses reasonable common cause combinations if in fact such combinations are necessary to result in a plant and non-routine system alignments.

trip.

A systematic approach to ensure all possible common cause combinations and routine and non-routine Recognize that given the plants asymmetries the CCF grouping is straightforward.

system alignments needs to be developed, and documented.

These results indicate that the impact of CCF (e.g., the preferred ac buses) is minimal when considering random failures. The PRA is satisfactory to achieve at least a Category II compliance with the ASME Standards IE A5 and A6.

Finding Resolved.

Page 28 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion IE-A6-01 Finding IE-A6-02 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text uncertainties.

When performing the Although it appears that an evaluation of CCFs was A detailed evaluation to address this finding was completed in Attachment systematic evaluation performed since IE_LOY1Q-Y20 and 3 of Reference [9]. In summary, the evaluation states:

required in IE-AS, INCLUDE IE_LO.,ALL4PREFAC, etc, were identified; however, A process to ensure all possible common cause combinations and routine initiating events resulting documentation of the systematic evaluation for the from multiple failures, if the elimination of other support system CCF events was and non-routine system alignments is theoretically achievable, but time consuming and inconsistent with risk-informed approaches as utilized in equipment failures result not provided.

PRAs. However, a process to ensure that combinations of failure events from a common cause, and Provide documentation of the evaluation of electrical and system configurations that have occurred or could reasonably occur is from routine system equipment CCF initiating events.

in place already through the current Palisades approach of considering alignments.

plant and generic data, initiating event categorization, and technical speCifications. This process addresses reasonable common cause combinations if in fact such combinations are necessary to result in a plant trip.

Finding Resolved.

When performing the Event trees for common cause failures (e.g., Loss of A detailed evaluation to address this finding was completed in Attachment systematic evaluation Preferred AC Bus Y20, Y30, and Y 40) are included in 3 of Reference [9].

required in IE-AS, INCLUDE the SAPHIRE program, but no documentation In summary, the evaluation states:

initiating events resulting associated with these event trees has been found.

from multiple failures, if the Note, the FMEA discussion provided in the Initiating A process to ensure all possible common cause combinations and routine equipment failures result Event notebook, does not specifically discuss the and non-routine system alignments is theoretically achievable, but time from a common cause, and CCF initiators, nor does it identify the buses as consuming and inconsistent with risk-informed approaches as utilized in from routine system necessarily resulting in a reactor trip. No discussion of PRAs. However, a process to ensure that combinations of failure events alignments.

non-routine system alignments has been found.

and system configurations that have occurred or could reasonably occur is The NRC's clarification for this element requires in place already through the current Palisades approach of considering consideration, and documentation of Initiating events plant and generic data, initiating event categorization, and technical resulting from common cause or from both routine specifications. This process addresses reasonable common cause combinations if in fact such combinations are necessary to result in a plant and non-routine system alignments.

trip.

A systematic approach to ensure all possible common Recognize that given the plants asymmetries the CCF grouping is cause combinations and routine and non-routine system alignments needs to be developed, and straightforward.

documented.

These results indicate that the impact of CCF (e.g., the preferred ac buses) is minimal when conSidering random failures. The PRA is satisfactory to achieve at least a Category II compliance with the ASME Standards IE AS and A6.

Finding Resolved.

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Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition (Supporting Suggestion Category II Text Requirement)

IE-A8-01 Finding INTERVIEW plant personal No meeting minutes or documentation of reviews Section 2.2 of NB-PSA-IE [10] was revised to document interviews and (e.g., operations, performed by Licensed operators, system engineers reviews of the PRA initiating events by specific plant personnel including maintenance, engineering, and maintenance and training staff members to the Assistant Operations Training Manager, Maintenance Rule Program and safety analysis) to ensure that no potential initiating events have been owner, and two operations personnel. In addition, the current Palisades determine if potential overlooked.

PRA personnel also act as the site safety analysis (Chapter 14) initiating events have been Lack of documentation of reviews performed by calculation owners. Interviews with System Engineers were performed by overlooked, the PRA personnel and documented in Attachment 5 of all PRA system Licensed operators, system engineers and notebooks. These interviews included discussion of initiating events.

maintenance and training staff members to ensure that potential initiating events have been overlooked.

Document review of IE List for comprehensiveness Finding Resolved.

performed by Licensed operators, system engineers and maintenance and training staff members.

IE-A9-01 Finding REVIEW plant-specific and Evaluation of precursors mentioned in Section 2.2.6 A documented review of all maintenance rule and work order failures was review industry operating Special Initiators as Special initiating events or the added to Section 2.2.6 of the initiating events notebook NB-PSA-IE [10] to experience for initiating potential for such events (e.g., precursors) was determine if they are potential precursor events. Component failures were event precursors, for performed during the PRA teams review of the obtained from Attachment 3 of the data notebook NB-PSA-DA [5] and identifying additional Maintenance Rule (MR) database and Maintenance individually evaluated as to their potential as a precursor event. No new initiating events. For Work Orders (MWO) in support of the data effort.

initiating events were developed as a result of the evaluation. However, example, plant-specific However, documentation of the specific review for the exercise did confirm several existing transient initiator events were experience with intake precursors was not provided, appropriately modeled in the PRA.

structure clogging might Provide documentation to show the evaluation Finding Resolved.

indicate that loss of intake structures should be performed.

identified as a potential initiating event.

IE-B3-01 Suggestion GROUP initiating events In grouping initiators with respect to plant impact, The addition of operator action timing was added to the initiating event only when the following is there was no explicit discussion of operator timing grouping criteria in the initiating events notebook NB-PSA-IE [10].

true: (a) events can be issues as they might impact the groupings.

The criteria added consider:

considered similar in terms of plant response, success Timing of operator actions may affect the accident criteria, timing, and the sequence progression to the extent they may be

1. plant response following the initiating event requires unique operator

actions, effect on the operability and sufficiently different to be considered in different performance of operators groups.

2. the initiating event disables instrumentation which is required for successful operator action, or and relevant mitigating Explicitly include consideration of operator action systems; or (b) events can timing in defining the initiator groups.

3. the initiating event changes the likelihood of successful operator be subsumed into a group performance by some other mechanism and bounded by the worst case impacts within the Suggestion Resolved Page 29 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion IE-A8-01 Finding IE-A9-01 Finding IE-B3-01 Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text INTERVIEW plant personal No meeting minutes or documentation of reviews Section 2.2 of NB-PSA-IE [10] was revised to document interviews and (e.g., operations, performed by Licensed operators, system engineers reviews of the PRA initiating events by specific plant personnel including maintenance, engineering, and maintenance and training staff members to the Assistant Operations Training Manager, Maintenance Rule Program and safety analysis) to ensure that no potential initiating events have been owner, and two operations personnel. In addition, the current Palisades determine if potential overlooked.

PRA personnel also act as the site safety analysis (Chapter 14) initiating events have been Lack of documentation of reviews performed by calculation owners. Interviews with System Engineers were performed by overlooked.

Licensed operators, system engineers and the PRA personnel and documented in Attachment 5 of all PRA system maintenance and training staff members to ensure notebooks. These interviews included discussion of initiating events.

that potential initiating events have been overlooked.

Document review of IE List for comprehensiveness Finding Resolved.

performed by Licensed operators, system engineers and maintenance and training staff members.

REVIEW plant-specific and Evaluation of precursors mentioned in Section 2.2.6 A documented review of all maintenance rule and work order failures was review industry operating Special Initiators as *Special initiating events or the added to Section 2.2.6 of the initiating events notebook NB-PSA-IE [1 OJ to experience for initiating potential for such events (e.g., precursors) was determine if they are potential precursor events. Component failures were event precursors, for performed during the PRA teams' review of the obtained from Attachment 3 of the data notebook NB-PSA-DA [5J and identifying additional Maintenance Rule (MR) database and Maintenance individually evaluated as to their potential as a precursor event. No new initiating events. For Work Orders (MWO) in support of the data effort.*

initiating events were developed as a result of the evaluation. However, example, plant-specific However, documentation of the specific review for the exercise did confirm several existing transient initiator events were experience with intake precursors was not provided.

appropriately modeled in the PRA.

structure clogging might Provide documentation to show the evaluation Finding Resolved.

indicate that loss of intake structures should be performed.

identified as a potential initiating event.

GROUP initiating events In grouping initiators with respect to plant impact, The addition of operator action timing was added to the initiating event only when the following is there was no explicit discussion of operator timing grouping criteria in the initiating events notebook NB-PSA-IE [10].

true: (a) events can be issues as they might impact the groupings.

The criteria added consider:

considered similar in terms Timing of operator actions may affect the accident of plant response, success sequence progression to the extent they may be

1. plant response following the initiating event requires unique operator criteria, timing, and the sufficiently different to be considered in different

actions, effect on the operability and performance of operators groups.

2. the initiating event disables instrumentation which is required for and relevant mitigating Explicitly include consideration of operator action successful operator action, or systems; or (b) events can timing in defining the initiator groups.

3. the initiating event changes the likelihood of successful operator be subsumed into a group performance by some other mechanism and bounded by the worst Suggestion Resolved case impacts within the Page 29 of 55

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Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) new group. DO NOT SUBSUME scenarios into a group unless (1) the impacts are comparable to or less than those of the remaining events in that group AND (2) it is demonstrated that such grouping does not impact significant accident sequences IE-B4-01 Suggestion GROUP separately from Palisades did not model excess LOCA such as Palisades was one of three pilot plants evaluated in the recent NRC effort other initiating event random vessel rupture based on their Pressurized to re-evaluate the nsk of pressurized thermal shock. These efforts are categories those categories Thermal Shock (PTS) evaluation. However, an summarized in NUREG-1 806 and NUREG-1874. The analyses made use with different plant response excessive LOCA event is explicitly called out in the of three Palisades specific analytical models (PRA, RELAP and FAVOR)

(i.e., those with different SR. However, because of the low generic Initiating that together, allowed the estimate of the yearly through-wall crack success rate criteria) event frequency, this is not expected to have a frequency (TWCF) in a reactor pressure vessel (RPV). Using the 20+

impacts or those that could significant impact on the results.

year old NURE-1150 data (the generic frequency) to model Excessive have more severe LOCAl Vessel Rupture in lieu of the latest plant specific state-of-Excessive LOCA/Vessel Rupture should be included knowledge based on the joint RES/Industry 50.61 initiative is not radionuclide release in the model as leading directly to core damage.

warranted. Note that the dominate sequence was a non-mechanistic potential (e.g., LERF). This Palisades can use the generic frequency or they can scenario that assumed the pressunzer safeties failed open for a period of includes such initiators as use the frequency from their Pressurized Thermal time and subsequently reclosed. The next set of dominant sequences did excessive LOCA, interfacing Shock Analysis.

not include a pressure component. Refer to NB-PSA-IE, Palisades systems LOCA, steam generator tube ruptures, and Probabilistic Safety Assessment Initiating Event Notebook, [10].

unisolated breaks outside NB-PSA-IE dedicates 4 pages addressing Pressurized Thermal Shock.

containment.

Palisades was one of three pilot plants evaluated in the NRC initiative to re-evaluate the risk of pressurized thermal shock. The analyses made use of three Palisades specific analytical models (PRA, RELAP and FAVOR) that together, allowed the estimate of the yearly through-wall crack frequency (TWCF) in the reactor pressure vessel (RPV).

Suggestion Resolved IE-Ci-Ol Suggestion CALCULATE the initiating This is in Reference to Section 3.1 of the Initiating Subsequent to performance of the peer review, NUREG/CR-7037, event frequency accounting Events Notebook (NB-PSA-lE): The thermal capacity Industry Performance of Relief Valves at U.S. Commercial Nuclear Power for relevant generic and of the steam generators at Palisades is such that a Plants through 2007, was published. This document provides an update plant-specific data unless it demand on the PORVs or pressurizer SRVs is not of the industry data related to safety valves and low capacity relief valves, is justified that there are expected following a reactor trip. This has been as well as illustrating an approach for modeling pressurizer safety valves, adequate plant-specific data validated per review of past thermal hydraulic main steam safety valves, atmospheric dump valves and PCS PORVs in to characterize the analyses (Final Safety Analysis Report (FSAR)

PWR PRAs.

parameter value and its Chapter 14). In addition, in the 30 plus years of Page 30 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion IE-94-01 Suggestion IE-C1-01 Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a{a){3){i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text "new" group. 00 NOT SUBSUME scenarios into a group unless (1) the impacts are comparable to or less than those of the remaining events in that group AND (2) it is demonstrated that such grouping does not impact significant accident sequences GROUP separately from Palisades did not model excess LOCA such as Palisades was one of three pilot plants evaluated in the recent NRC effort other initiating event random vessel rupture based on their Pressurized to re-evaluate the risk of pressurized thermal shock. These efforts are categories those categories Thermal Shock (PTS) evaluation. However, an summarized in NUREG-1806 and NUREG-1874. The analyses made use with different plant response excessive LOCA event is explicitly called out in the of three Palisades specific analytical models (PRA, RELAP and FAVOR)

(i.e., those with different SA. However, because of the low generic Initiating that together, allowed the estimate of the yearly through-wall crack success rate criteria) event frequency, this is not expected to have a frequency (TWCF) in a reactor pressure vessel (RPV). Using the 20+

impacts or those that could significant impact on the results.

year old NURE-1150 data (athe generic frequency*) to model Excessive have more severe Excessive LOCANessel Rupture should be included LOCAl Vessel Rupture in lieu of the latest plant specific state-of-radionuclide release in the model as leading directly to core damage.

knowledge based on the joint RES/Industry 50.61 initiative is not potential (e.g., LERF). This Palisades can use the generic frequency or they can warranted. Note that the dominate sequence was a non-mechanistic includes such initiators as use the frequency from their Pressurized Thermal scenario that assumed the pressurizer safeties failed open for a period of excessive LOCA, interfacing Shock Analysis.

time and subsequently reclosed. The next set of dominant sequences did systems LOCA, steam not include a pressure component. Refer to NB-PSA-IE, "Palisades generator tube ruptures, and Probabilistic Safety Assessment Initiating Event Notebook", [10].

unisolated breaks outside NB-PSA-IE dedicates 4 pages addreSSing Pressurized Thermal Shock.

containment.

Palisades was one of three pilot plants evaluated in the NRC initiative to re-evaluate the risk of pressurized thermal shock. The analyses made use of three Palisades specific analytical models (PRA, RELAP and FAVOR) that together, allowed the estimate of the yearly through-wall crack frequency (lWCF) in the reactor pressure vessel (RPV).

Suggestion Resolved CALCULATE the initiating This is in Reference to Section 3.1 of the Initiating Subsequent to performance of the peer review, NUREG/CR-7037, event frequency accounting Events Notebook (NB-PSA-IE): aThe thermal capacity "Industry Performance of Relief Valves at U.S. Commercial Nuclear Power for relevant generic and of the steam generators at Palisades is such that a Plants through 2007," was published. This document provides an update plant-specific data unless it demand on the PORVs or pressurizer SRVs is not of the industry data related to safety valves and low capacity relief valves, is justified that there are expected following a reactor trip. This has been as well as illustrating an approach for modeling pressurizer safety valves, adequate plant-specific data validated per review of past thermal hydraulic main steam safety valves, atmospheric dump valves and PCS PORVs in to characterize the analyses (Final Safety Analysis Report (FSAR)

PWR PRAs.

parameter value and its Chapter 14). In addition, in the 30 plus years of Page 30 of 55

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Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) uncertainty. (See also IE-operation, the plant has not experienced such an Based on that document, Palisades has revised its approach to modeling C13 for requirements for event. Moreover, the Palisades nominal operating consequential and spurious pressurizer safety valve demands [29].

rare and extremely rare pressure of 2060 psia is about 100 psi less than that Pressurizer safety valve opening is no longer considered an initiating events),

of all PWRs. Only inadvertent or premature operation event, but is considered as a consequential or spunous event.

of these valves can lead to loss of coolant type Suggestion Resolved conditions. Given a demand and subsequent failure of the pressurizer SRVs, the consequences of a small break LOCA are analyzed by linking to a replication of the baseline small break LOCA event tree. Although Palisades has not experienced pressurizer safety valve setup or setpoint drift problems, operating experience [e.g., Fort Calhoun (Licensee Event Report (LER) 285/92-028) and Calvert Cliffs (LER 31 7/94-007)] has shown that such events are plausible. As such, this event has been included in the model (EA-PSA-PSAR2-04-02).

Ensure that the Palisades definition for IE-LOCA PZRSRV is consistent with the definition and events used to calculate NUREG/CR-6928s for lE-SORV (PWR). The events in NUREG/CR-i6928 lE-SORV were used directly in defining the prior, but are actually consequential SORV following another initiating event versus a spurious opening of a relief valve.

Documentation could be improved.

lE-C2-01 Finding When using plant-specific Justification for the exclusion of data before January Added additional justification for the exclusion of data prior to January data, USE the most recent 2003 used to identify plant-specific initiating events 2003 to Section 4.1 of the initiating events notebook NB-PSA-IE [10].

applicable data to quantify was not provided.

Justification is based on improved plant availability from January 2003

the initiating event Justification for the exclusion of data before January 2009 relative to the previous site specific initiating event data from January 1994 December 2002. Improvements in plant availability were frequencies. JUSTIFY 2003 used to identify plant-specific initiating events demonstrated graphically in Figure 4.1. Plant availability has excluded data that is not was not provided, demonstrably improved after January 2003 due improved operating and considered to be either recent or applicable (e.g.,

Provide the requested justification.

maintenance practices.

provide evidence via design Finding Resolved.

or operational change that the data are no longer applicable.)

lE-C6-01 Finding USE as screening criteria no In relation to lE-C6, Operator actions are apparently The basis for excluding control room HVAC from the full power internal higher than the following credited for the exclusion of some events (e.g.,

events model was strengthened to include other aspects in addition to Page 31 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion IE-C2-01 Finding IE-C6-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED AL~ERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text uncertainty. (See also IE-operation, the plant has not experienced such an Based on that document, Palisades has revised its approach to modeling C13 for requirements for event. Moreover, the Palisades nominal operating consequential and spurious pressurizer safety valve demands [29].

rare and extremely rare pressure of 2060 psia is about 100 psi less than that Pressurizer safety valve opening is no longer considered an initiating events).

of all PWRs. Only inadvertent or premature operation event, but is considered as a consequential or spurious event.

of these valves can lead to loss of coolant type Suggestion Resolved conditions. Given a demand and subsequent failure of the pressurizer SRV's, the consequences of a small break LOCA are analyzed by linking to a replication of the baseline small break LOCA event tree. Although Palisades has not experienced pressurizer safety valve setup or setpoint drift problems, operating experience [e.g., Fort Calhoun (Licensee Event Report (LER) 285/92-028) and Calvert Cliffs (LER 317/94-007)] has shown that such events are plausible. As such, this event has been included in the model (EA-PSA-PSAR2-04-02). A Ensure that the Palisades definition for IE-LOCA-PZRSRV is consistent with the definition and events used to calculate NUREGlCR-6928's for IE-SORV (PWR). The events in NUREGlCR...,6928IE-SORV were used directly in defining the prior, but are actually consequential SORV following another initiating event versus a spurious opening of a relief valve.

Documentation could be improved.

When using plant-specific Justification for the exclusion of data before January Added additional justification for the exclusion of data prior to January data, USE the most recent 2003 used to identify plant-specific initiating events 2003 to Section 4.1 of the initiating events notebook NB-PSA-IE [10].

applicable data to quantify was not provided.

Justification is based on improved plant availability from January 2003-the initiating event Justification for the exclusion of data before January 2009 relative to the previous site specific initiating event data from frequencies. JUSTIFY 2003 used to identify plant-specific initiating events January 1994 - December 2002. Improvements in plant availability were excluded data that is not demonstrated graphically in Figure 4.1. Plant availability has considered to be either was not provided.

demonstrably improved after January 2003 due improved operating and recent or applicable (e.g.,

Provide the requested justification.

maintenance practices.

provide evidence via design Finding Resolved.

or operational change that the data are no longer applicable.)

USE as screening criteria no In relation to IE-C6, Operator actions are apparently The basis for excluding control room HVAC from the full power internal higher than the following credited for the exclusion of some events (e.g.,

events model was strengthened to include other aspects in addition to Page 31 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) characteristics (or more stringent characteristics as devised by the analyst) to eliminate initiating events or groups from further evaluation: (a) the frequency of the event is less than 1 E 7 per reactor year (/ry), and the event does not involve either an ISLOCA, containment bypass, or reactor pressure vessel rupture (b) the frequency of the event is less than 1 E 6/ry, and core damage could not occur unless at least two trains of mitigating systems are failed independent of the initiator, or (c) the resulting reactor shutdown is not an immediate occurrence. That is, the event does not require the plant to go to shutdown conditions until sufficient time has expired during which the initiating event conditions, with a high degree of certainty (based on supporting calculations),

are detected and corrected before normal plant operation is curtailed (either administratively or automatically). If either criterion (a) or (b) above is used, then CONFIRM that the value specified in the criterion meets the applicable requirements in Data Analysis (2-2.6) and Level 1 Quantification (2-2.7).

CRHVAC refer to earlier HVAC comments) without justifying each such credit (operator training, procedures, etc.)

If component/system failures lead to an initiating event but are screened from further analysis by crediting operator actions or equipment/systems to avert the transient, then quantify the total initiating event frequency considering these events and apply criteria of IE-C6 to determine if screening criteria is met.

Apply IE-C6 screening criteria and document as appropriate.

operator actions and was fully documented in Attachment 8 of NB-PSA ETSC [11]. The evaluation was updated to include discussion of the control room heat-up rate effects on the reactor protective system (RPS) components and concluded that a loss of HVAC would not result in a significant increase in the failure probability of the RPS.

In addition, a comparison of sensitivity analyses performed based on 14 owners group sites that modeled the contribution to CDF due to loss of control room HVAC. The sensitivity studies found that the average CDF/yr was 1.61 E-07 with a median of 1.31 E-07/yr. Given Palisades core damage frequency is on the order of E-05, the change in CDF due to loss of control room HVAC would less than 1%.

With respect to cable spreading room cooling. An analysis of the cable spreading room heat-up following a loss of ventilation was developed using the GOTHIC software code and documented in EA-PSA-GOTHIC CSRHEATUP-09-09 Rev. 0 [12]. This analysis developed a conservative room heat-up profile based on actual test data and assuming operators take no action to either open doors or affix portable ventilation. Using the room heat-up profile output from the analysis, CALC-455-001 -DC2 [13]

was then performed to evaluate all cable spreading room equipment modeled in the PRA at the predicted peak temperature for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Based on the evaluation of equipment qualification reports, and vendor data, it was concluded there is reasonable assurance of operability for all equipment in the room under these conditions.

The conclusions of these analyses demonstrate ventilation to the cable spreading and control room areas is not necessary to be explicitly modeled and the bases for these conclusions do not require operator action to mitigate elevated temperatures. However, ventilation is considered for purposes of fire modeling in these areas.

Finding Resolved.

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition (Supporting Suggestion Category II Text Requirement)

Page 32 Of 55 F&O#

Finding or (Supporting Requirement)

Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text characteristics (or more CRHVAC refer to earlier HVAC comments) without operator actions and was fully documented in Attachment 8 of NB-PSA-stringent characteristics as justifying each such credit (operator training, ETSC [11]. The evaluation was updated to include discussion of the devised by the analyst) to procedures, etc.)

control room heat-up rate effects on the reactor protective system (RPS) eliminate initiating events or If component/system failures lead to an initiating components and concluded that a loss of HVAC would not result in a groups from further significant increase in the failure probability of the RPS.

evaluation: (a) the frequency event but are screened from further analysis by of the event is less than 1 E-crediting operator actions or equipment/systems to In addition, a comparison of sensitivity analyses performed based on 14 avert the transient, then quantify the total initiating owner's group sites that modeled the contribution to CDF due to loss of 7 per reactor year (fry), and the event does not involve event frequency considering these events and apply control room HVAC. The sensitivity studies found that the average CDFfyr either an ISLOCA, criteria of IE-C6 to determine if screening criteria is was 1.61 E-07 with a median of 1.31 E-07fyr. Given Palisades core containment bypass, or met.

damage frequency is on the order of E-05, the change in CDF due to loss reactor pressure vessel Apply IE-C6 screening criteria and document as of control room HVAC would less than 1 %.

rupture (b) the frequency of appropriate.

With respect to cable spreading room cooling. An analysis of the cable I

I the event is less than 1 E-spreading room heat-up following a loss of ventilation was developed 6fry, and core damage could using the GOTHIC software code and documented in EA-PSA-GOTHIC-not occur unless at least two CSRHEATUP-09-09 Rev. 0 [12]. This analysis developed a conservative trains of mitigating systems room heat-up profile based on actual test data and assuming operators are failed independent of the take no action to either open doors or affix portable ventilation. Using the initiator, or (c) the resulting room heat-up profile output from the analysis, CALC-455-001-DC2 [13]

reactor shutdown is not an was then performed to evaluate all cable spreading room equipment immediate occurrence. That modeled in the PRA at the predicted peak temperature for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

is, the event does not Based on the evaluation of equipment qualification reports, and vendor require the plant to go to data, it was concluded there is reasonable assurance of operability for all shutdown conditions until equipment in the room under these conditions.

sufficient time has expired The conclusions of these analyses demonstrate ventilation to the cable during which the initiating event conditions, with a high spreading and control room areas is not necessary to be explicitly degree of certainty (based modeled and the bases for these conclusions do not require operator on supporting calculations),

action to mitigate elevated temperatures. However, ventilation is are detected and corrected considered for purposes of fire modeling in these areas.

before normal plant Finding Resolved.

operation is curtailed (either administratively or automatically). If either criterion (a) or (b) above is used, then CONFIRM that the value specified in the criterion meets the applicable requirements in Data AnalysiS (2-2.6) and Level 1 Quantification (2-2.7).

Page 32 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition (Supporting Suggestion Category II Text Requirement)

IE-C12-01 Suggestion COMPARE results and Section 5 of NB-PSA-lErl describes the quantification Table 5.15 was added to Section 5.12 of the initiating events notebook EXPLAIN differences in the of the Initiating Event Frequencies. As part of this NB-PSA-IE [10]. This table presents a comparison of Palisades initiating initiating event analysis with quantification, data from numerous outside sources events and frequencies to those developed at Waterford 3, which is a generic data sources to and is combined with plant specific data via Bayesian similar Combustion Engineering designed PWR. Where significant provide a reasonableness updates. However, there was no comparison of the differences are noted the table provides additional notes.

check of the results.

results to the frequencies used by other similar plants.

In addition, LOCA IE frequency validation occurred during conduct of the The SR requires a reasonableness check of the Palisades pressurized thermal shock (PTS) analyses. There was a initiating event frequencies against those of other concern that the LOCA frequencies in NUREG/CR-5750 did not account plants. While Palisades did do comparisons against for age-related factors important to deriving the frequencies. An expert generic data, there was no plant to plant comparison elicitation effort (independent of RES) at the NRC was conducted to in most cases.

account for these adjustments. The NRC expert elicitation subject matter experts concluded that the Palisades plant specific initiating event Palisades should include a table showing their frequencies (employed in the 50.61 RES / Industry initiative and used in initiating event frequencies and the equivalent the current internal events analysis) were nearly the same as that frequencies for one or more plants of similar vintage, developed in the elicitation effort.

Where there are large differences, Palisades should explain and justify the differences.

Palisades now utilizes LOCA IE frequencies based on NUREG-1 829.

Suggestion resolved.

lE-C14-01 Suggestion In the ISLOCA frequency The Palisades Interfacing Systems LOCA (ISLOCA)

ISLOCA modeling was updated in Attachment F of analysis, INCLUDE the models considered the failure of the first check valve, EA-PSA-FPIE-FIRE-12-04. The fault trees now consider potential following features of plant i.e., CK-ES31 01, CK-E531 16, CK-ES31 31 and CK-ISLOCAs occurring through all primary coolant system interfaces with the and procedures that ES-3146, as the initiator for the ISLOCA sequences.

high pressure safety injection, low pressure safety injection, shutdown influence the ISLOCA Palisades does not consider failure of one of the other cooling, and charging systems. The model now considers potential frequency: (a) configuration valves as potential initiators because of a pair of ISLOCA through 5 containment penetrations with 17 potential flow paths.

of potential pathways quarterly tests, one of which demonstrates the valves including numbers and types will open and one that confirms that the valves Suggestion Resolved of valves and their relevant reclose. The standby failure rate for one quarter is failure modes and the used to calculate the failure probability for these existence, size, and valves fail open. The fact that these tests were positioning of relief valves sequenced such that the valves were confirmed (b) provision of protective closed on a quarterly basis was not immediately interlocks (c) relevant apparent in the documentation.

surveillance test procedures Palisades should revise their ISLOCA analysis (d) the capability of documentation to clearly demonstrate the fact that secondary system piping (e) these tests were sequenced such that the valves isolation capabilities given were confirmed closed on a quarterly basis. This was high flow/differential not immediately apparent in the documentation.

pressure conditions that might_exist_following_breach Page 33 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion IE-C12-01 Suggestion IE-C14-01 Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text COMPARE results and Section 5 of NB-PSA-IEr1 describes the quantification Table 5.15 was added to Section 5.12 of the initiating events notebook EXPLAIN differences in the of the Initiating Event Frequencies. As part of this NB-PSA-IE [10]. This table presents a comparison of Palisades initiating initiating event analysis with quantification, data from numerous outside sources events and frequencies to those developed at Waterford 3, which is a generic data sources to and is combined with plant specific data via Bayesian similar Combustion Engineering designed PWR. Where significant provide a reasonableness updates. However, there was no comparison of the differences are noted the table provides additional notes.

check of the results.

results to the frequencies used by other similar plants.

In addition, LOCA IE frequency validation occurred during conduct of the The SR requires a reasonableness check of the Palisades pressurized thermal shock (PTS) analyses. There was a initiating event frequencies against those of other concem that the LOCA frequencies in NUREGlCR-5750 did not account plants. While Palisades did do comparisons against for age-related factors important to deriving the frequencies. An expert generic data, there was no plant to plant comparison elicitation effort (independent of RES) at the NRC was conducted to in most cases.

account for these adjustments. The NRC expert elicitation subject matter Palisades should include a table showing their experts concluded that the Palisades plant specific initiating event frequencies (employed in the 50.61 RES / Industry initiative and used in initiating event frequencies and the equivalent the current intemal events analysis) were nearly the same as that frequencies for one or more plants of similar vintage.

developed in the elicitation effort.

Where there are large differences, Palisades should explain and justify the differences.

Palisades now utilizes LOCA IE frequencies based on NUREG-1829.

Suggestion resolved.

In the ISLOCA frequency The Palisades Interfacing Systems LOCA (ISLOCA)

ISLOCA modeling was updated in Attachment F of analysis, INCLUDE the models considered the failure of the first check valve, EA-PSA-FPIE-FIRE-12-04. The fault trees now consider potential following features of plant Le., CK-ES3101, CK-ES3116, CK-ES3131and CK-ISLOCAs occurring through all primary coolant system interfaces with the and procedures that ES-3146, as the initiator for the ISLOCA sequences.

high pressure safety injection, low pressure safety injection, shutdown influence the ISLOCA Palisades does not consider failure of one of the other cooling, and charging systems. The model now considers potential frequency: (a) configuration valves as potential initiators because of a pair of ISLOCA through 5 containment penetrations with 17 potential flow paths.

of potential pathways quarterly tests, one of which demonstrates the valves Suggestion Resolved including numbers and types will open and one that confirms that the valves of valves and their relevant reclose. The standby failure rate for one quarter is failure modes and the used to calculate the failure probability for these existence, size, and valves fail open. The fact that these tests were positioning of relief valves sequenced such that the valves were confirmed (b) provision of protective closed on a quarterly basis was not immediately interlocks (c) relevant apparent in the documentation.

surveillance test procedures Palisades should revise their ISLOCA analysis (d) the capability of documentation to clearly demonstrate the fact that secondary system piping (e) these tests were sequenced such that the valves isolation capabilities given were confirmed closed on a quarterly basis. This was high flow/differential not immediately apparent in the documentation.

pressure conditions that might exist following breach Page 33 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions JUSTIFY any credit given for equipment survivability or human actions under adverse environments.

No credit is taken for equipment operability or operator actions for adverse environment or containment failure. In Section 6.2.4 of the Level 2 report, Palisades stated that they had reviewed the results for cases where credit for equipment or HRAs during harsh environment or after containment failure might be applicable but did not justify equipment survivability in either of these conditions based on the contention that there were no cases where crediting continued equipment operation or operator actions would affect LERF. Therefore no credit was taken for continued equipment operation or operator actions.

This clearly meets the requirements for Capability Category I To move up to Capability Category Il/Ill, i.e., getting credit for not crediting equipment, Palisades would need to provide much more documentation on what was looked at for equipment operability or operator actions and provide the bases for why the equipment would not be operable or that crediting the equipment made no difference to LERF. This should be tied to the Severe Accident Mitigation Guidelines (SAMG)

No credit is taken for equipment operability or operator actions in adverse environments or after containment failure. Palisades reviewed the LERF results for opportunities to take such credit (as documented in Section 6.2.4 of the Level 2 Notebook) and justified the lack of credit.

Based on way the standard is written, the only way to earn a CC-Il categorization is to credit equipment operation in adverse environment (for LE-C9 and C-i 0) and after containment failure (for LE-Ci 1 and Cl 2).

Moreover, from an equipment context, Palisades does credit equipment in containment in environments that are considered beyond the EEQ harsh environment for which the equipment is qualified in the design basis.

The MAAP program was utilized in calculation PLP0247-07-0004.0i R2

[14] to determine the bounding best-estimate containment environmental conditions postulated to be encountered by equipment located in containment and modeled in the PRA. Both single and double steam generator blowdowns inside containment as well as once-through-cooling events were analyzed, with either a single containment air cooler or a single containment spray pump and spray header available. Additional variations with respect to steam generator isolation and auxiliary feedwater flow were analyzed. The limiting conditions are considered to represent the worst containment conditions expected prior to core damage and vessel failure, and are clearly beyond the design basis of the plant given the assumption of a double steam generator blowdown and that only portions of redundant containment heat removal systems available.

Calculation CALC-455-00i-DC1 [15] evaluates the survivability of equipment modeled in the PRA under the environmental conditions determined in the MAAP analyses. This analysis utilized temperature profiles from the MAAP program to demonstrate that all credited PRA equipment located in containment can survive the limiting containment conditions produced by MSLB, LOCA, and OTC scenarios in which only a single containment air cooler or a single containment spray pump and header are available. A further detailed summary is provided in of NB-PSA-ETSC [11].

In summary, it is considered that supporting requirements LE-C9 and LE-ClO meet the CC-Il requirements as the above noted engineering evaluations provide the justification.

Palisades does not take credit for continued operation of equipment or operator action after containment failure. Therefore, by definition, a CC-I LE-C9-01 Suggestion F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Requirement)

Suggestion Category II Text Finding Description (summary discussion)

Disposition of the secondary system Page 34 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion LE-C9-01 Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text of the secondary system JUSTIFY any credit given No credit is taken for equipment operability or No credit is taken for equipment operability or operator actions in adverse for equipment survivability or operator actions for adverse environment or environments or after containment failure. Palisades reviewed the LERF human actions under containment failure. In Section 6.2.4 of the Level 2 results for opportunities to take such credit (as documented in Section adverse environments.

report, Palisades stated that they had reviewed the 6.2.4 of the Level 2 Notebook) and justified the lack of credit.

results for cases where credit for equipment or HAAs Based on way the standard is written, the only way to eam a CC-II during harsh environment or after containment failure might be applicable but did not justify equipment categorization is to credit equipment operation in adverse environment (for survivability in either of these conditions based on the LE-C9 and C-10) and after containment failure (for LE-C11 and C12).

contention that there were no cases where crediting Moreover, from an equipment context, Palisades does credit equipment in continued equipment operation or operator actions containment in environments that are considered beyond the EEQ harsh would affect LERF. Therefore no credit was taken for environment for which the equipment is qualified in the design basis.

continued equipment operation or operator actions.

The MAAP program was utilized in calculation PLP0247-07-0004.01R2 This clearly meets the requirements for Capability Category I

[14] to determine the bounding best-estimate containment environmental conditions postulated to be encountered by equipment located in To move up to Capability Category 11/111, i.e., getting containment and modeled in the PRA. Both single and double steam credit for not crediting equipment, Palisades would generator blowdowns inside containment as well as once-through-cooling need to provide much more documentation on what events were analyzed, with either a Single containment air cooler or a was looked at for equipment operability or operator single containment spray pump and spray header available. Additional actions and provide the bases for why the equipment variations with respect to steam generator isolation and auxiliary would not be operable or that crediting the equipment feedwater flow were analyzed. The limiting conditions are considered to made no difference to LERF. This should be tied to represent the worst containment conditions expected prior to core damage the Severe Accident Mitigation Guidelines (SAMG) and vessel failure, and are clearly beyond the design basis of the plant given the assumption of a double steam generator blowdown and that only portions of redundant containment heat removal systems available.

Calculation CALC-455-001-DC1 [15] evaluates the survivability of equipment modeled in the PRA under the environmental conditions determined in the MAAP analyses. This analysis utilized temperature profiles from the MAAP program to demonstrate that all credited PAA equipment located in containment can survive the limiting containment conditions produced by MSLB, LOCA, and OTC scenarios in which only a single containment air cooler or a Single containment spray pump and header are available. A further detailed summary is provided in of NB-PSA-ETSC [11].

In summary, it is considered that supporting requirements LE-C9 and LE-C10 meet the CC-II requirements as the above noted engineering evaluations provide the justification.

Palisades does not take credit for continued operation of equipment or operator action after containment failure. Therefore, by definition, a CC-I Page 34 of 55 I

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

F d

ASME R G *d 1 200 (Supporting Suggestkn Category II Text e

Finding Description (summary discussion)

Disposition Requirement) classification of supporting requirements LE-Ci 1 and Cl 2 is considered appropriate.

Finding Resolved LE-G5-01 Finding IDENTIFY limitations in the The Palisades PSA Level 2 Notebook does not Given the Palisades two source term models, PAL-L2 and PWROG-L2, it LERF analysis that would explicitly discuss any limitations in the LERF analysis is considered that sufficient detail exists such that this requirement is met.

impact applications, that might impact applications.

However, consideration of developing guidance as to when to use both models will be evaluated [26].

It is expected that the limitations will be similar to those discussed for the level 1 analyses, but the level Finding Resolved 1 discussion does not explicitly cover LERF so their analysis does not comply with the SR.

Palisades should develop such a discussion similar to that developed for the level 1 analyses or revise the level 1 discussion to include LERF.

MU-B2-01 Finding changes that would impact The Palisades analysis of record is PSAR2c. The The Palisades analysis of record is PSAR2c. The version release risk-informed decisions model was revised to include modifications needed presented to the peer review team on 10/26/10 was PSAR3 Release 2b.

should be prioritized to for the internal flooding analysis. The current release The PSAR3 release series is a set of updates that address Reg. Guide ensure that the most presented to the peer review team is PSAR3 Release 1.200, as well as a variety of NFPA-805 issues from multiple spurious significant changes are 2b. This model contains the updates associated with operations to spurious containment high pressure. The update process incorporated as soon as the requirements of Reg Guide 1.200 as well as was described during the peer review 10/26/2009 Monday morning practical.

changes to address NFPA-805. PSAR3 Release 2b is introduction. At the time of the peer review these releases are not not the current analysis of record.

validated analyses-of-record.

The purpose of providing the PSAR3 Release 2b results to the peer review team was to show the latest consequences from a variety of model updates ranging from component random failure data, lE frequencies etc.,

updating to addressing extensive flow diversion scenarios, to adaptation of the simplified Westinghouse LERF model, to incorporation of a new comprehensive common cause model that employed the latest data, etc.

PSAR3 Release 2b differed from PSAR3 Release 2a due to inclusion of new IE, HRA data etc.

With exception of the last significant plant modification (GSI-191 sump strainers) that was finalized in the spring of 2009, all significant modifications had been addressed in the current analysis-of-record, the PSAR2c model dated 6-30-2006. Both PSAR3 Release 2# series included the finalized GSI-191 modifications. These GSI-191 physical modifications were completed in the spring of 2009 during the scheduled REFOUT. These modifications included an extensive re-analysis of the passive_screen_design_due_to_the_reconciled_chemical-effect_tests_that Page 35 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion LE-GS-01 Finding MU-B2-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text classification of supporting requirements LE-C11 and C12 is considered appropriate.

Finding Resolved IDENTIFY limitations in the The Palisades PSA Level 2 Notebook does not Given the Palisades two source term models, PAL-L2 and PWROG-L2, it LERF analysis that would explicitly discuss any limitations in the LERF analysis is considered that sufficient detail exists such that this requirement is met.

impact applications.

that might impact applications.

However, consideration of developing guidance as to when to use both It is expected that the limitations will be similar to models will be evaluated [26].

those discussed for the level 1 analyses, but the level Finding Resolved 1 discussion does not explicitly cover LERF so their analysis does not comply with the SA.

Palisades should develop such a discussion similar to that developed for the level 1 analyses or revise the level 1 discussion to include LERF.

Changes that would impact The Palisades analysis of record is PSAR2c. The The Palisades analysis of record is PSAR2c. The version release risk-informed decisions model was revised to include modifications needed presented to the peer review team on 10/26/10 was "PSAR3 Release 2b".

should be prioritized to for the intemal flooding analysis. The current release The PSAR3 release series is a set of updates that address Reg. Guide ensure that the most presented to the peer review team is PSAR3 Release 1.200, as well as a variety of NFPA-80S issues from multiple spurious Significant changes are 2b. This model contains the updates associated with operations to spurious containment high pressure. The update process incorporated as soon as the requirements of Reg Guide 1.200 as well as was described during the peer review 10/26/2009 Monday moming practical.

changes to address NFPA-80S. PSAR3 Release 2b is introduction. At the time of the peer review these releases are not not the current analysis of record.

validated analyses-of-record.

The purpose of providing the PSAR3 Release 2b results to the peer review team was to show the latest consequences from a variety of model updates ranging from component random failure data, IE frequencies etc.,

updating to addressing extensive flow diversion scenarios, to adaptation of the simplified Westinghouse LERF model, to incorporation of a new comprehensive common cause model that employed the latest data, etc.

PSAR3 Release 2b differed from PSAR3 Release 2a due to inclusion of new IE, HRA data etc.

With exception of the last Significant plant modification (GSI-191 sump strainers) that was finalized in the spring of 2009, all significant modifications had been addressed in the current analysis-of-record, the PSAR2c model dated 6-30-2006. Both PSAR3 Release 2# series included the finalized GSI-191 modifications. These GSI-191 physical modifications were completed in the spring of 2009 during the scheduled REFOUT. These modifications included an extensive re-analysis of the passive screen design due to the reconciled chemical-effect tests that Page 35 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions p#orting Finding or ASMEReg.Gtde 1.200 Finding Description (summary discussion)

Disposition Requirement) occurred in the fall of 2008.

Given the PRA teams design basis responsibilities, support of this initiative was very involved and included support of an onsite NRC inspection that was required for startup approval. The GSI-191 mod was the most significant plant change that had occurred since release of the current analysis-of-record dated 6/30/2006 (PSAR2c).

Release 2c has since been completed and the Palisades QA process applied in establishing the new analysis-of-record termed PSAR3 to support the NFPA-805 initiative. This model version is used as the base model to support the flooding PRA and the interim FPIE PSAR 3.3.0.

The PRA model is a living analysis. The configuration management procedures are applied to control, develop, and adapt parallel models.

This is not considered a finding.

The GSI-1 91 modification addressed the uncertainty associated with LOCA generated debris and its impact on the plants recirculation actuation system. The reliability of the original plant sump strainers is considered to have improved given the addition of several orders of magnitude of additional screen surface area.

This issue is resolved as PSAR3 is now the analysis of record.

MU-B3-01 Suggestion PRA changes shall be Section 6.2 of the Configuration Control Notebook Sections 3.3 and 6.2 of the configuration control notebook (NB-PSA-CC performed consistent with requires review of model revisions to ensure that they

[17]) have been revised to include a requirement for the review of updates the previously defined appropriately implemented.

and upgrades against the ASME standard.

Supporting Requirements.

The configuration control document does not Suggestion Resolved specifically indicate that updates are to be done in accordance with corresponding SRs from the standard, but it is assumed that the definition of appropriately implemented includes such as review because the associated system, IE, or other notebooks that would be updated all currently have a section for self-assessment against the standard.

Add a sentence to the configuration control document to clarify that appropriately implemented means conformance to the standard supporting requirements.

MU-B4-01 Finding PRA Upgrades shall receive The Configuration Control Notebook specifies the Section 3.3 of the configuration control notebook NB-PSA-CC [17] has a peer review (in difference between and update and an upgrade but been revised to include a requirement for a peer review against the ASME Page 36 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion MU-B3-01 Suggestion MU-84-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text occurred in the fall of 2008.

Given the PRA teams design basis responsibilities, support of this initiative was very involved and included support of an on site NRC inspection that was required for startup approval. The GSI-191 mod was the most significant plant change that had occurred since release of the current analysis-of-record dated 6/30/2006 (PSAR2c).

Release 2c has since been completed and the Palisades QA process applied in establishing the anew analysis-of-recorda termed PSAR3 to support the NFPA-805 initiative. This model version is used as the base model to support the flooding PRA and the interim FPIE PSAR 3.3.0.

The PRA model is a living analysis. The configuration management procedures are applied to control, develop, and adapt parallel models.

This is not considered a finding.

The GSI-191 modification addressed the uncertainty associated with LOCA generated debris and its impact on the plants recirculation actuation system. The reliability of the original plant sump strainers is considered to have improved given the addition of several orders of magnitude of additional screen surface area.

This issue is resolved as PSAR3 is now the analysis of record.

PRA changes shall be Section 6.2 of the Configuration Control Notebook Sections 3.3 and 6.2 of the configuration control notebook (NB-PSA-CC performed consistent with requires review of model revisions to ensure that they

[17]) have been revised to include a requirement for the review of updates the previously defined appropriately implemented.

and upgrades against the ASME standard.

Supporting Requirements.

The configuration control document does not Suggestion Resolved specifically indicate that updates are to be done in accordance with corresponding SRs from the standard, but it is assumed that the definition of

-appropriately implemented" includes such as review because the associated system, IE, or other notebooks that would be updated all currently have a section for self-assessment against the standard.

Add a sentence to the configuration control document to clarify that -appropriately implemented" means conformance to the standard supporting requirements.

PRA Upgrades shall receive The Configuration Control Notebook specifies the Section 3.3 of the configuration control notebook NB-PSA-CC [17] has a peer review (in differen~ bet\\Y~n and update and an upgrade but been revised to include a requirement for a peer review against the ASME Page 36 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) accordance with the does not specifically require performance of a peer standard for PSA model upgrades.

requirements specified in review for upgrades.

Section 6 of the ASME PRA Finding Resolved The Standard specifically calls for a peer review for Standard) for those aspects of the PRA that have been PRA Upgrades, but the Configuration Control Notebook does not specifically call for one following upgraded. Refer to Section 2 of the ASME PRA an Upgrade.

Standard for the distinction Modify the Configuration Control Notebook to specify of a PRA Upgrade versus that peer reviews are required for PRA Upgrades.

PRA maintenance and update.

MU-Di-Ol Finding The PRA configuration The Configuration Control Notebook does not direct Section 3.3 of the configuration control notebook NB-PSA-CC [17] has control process shall include that updates or upgrades are compared with previous been revised to include requirements for the review of updates and evaluation of the impact of risk-informed decisions and have used the PRA.

upgrades against previous applications and analyses.

changes on previously implemented risk-informed Review of previous RI applications is not called out in Finding Resolved decisions that have used the the Configuration Control Notebook.

PRA AND that affect the Add requirement for reviewing the previous RI safe operation of the plant.

applications against the new PRA results to see if they impact the results of the previous work.

QU-Al-Ol Suggestion INTEGRATE the accident Figure 5-1 of the Quantification Report provides a To be implemented in a future notebook revision. This is a documentation sequences, system models, small flow chart on the process of integrating the issue only. No impact to the interim FPIE application for RI-ISI screening.

data, and HRA in the CAFTA models into the SAPHIRE code and the Suggestion Open quantification process for additional APIs used to prepare the SAPHIRE model each initiating event group, for quantification (including the integration of CCF accounting for system trees, HRA rules, etc.) While this flow chart gives an dependencies, to arrive at upper level explanation of the process, a more accident sequence detailed flow chart would be useful in ensuring a frequencies.

consistent integration for personnel that do not perform this task frequently.

The integration process is fairly complex and involves multiple codes and tools. Missing any step in this process could impact quantification.

Develop a more detailed flow chart for those performing the quantification.

QU-A3-01 Finding ESTIMATE the mean CDF The mean ISLOCA CDF frequency does not account A method of demonstrating the effect of the state of knowledge is to accounting for the state-of-for the state-of-knowledge correlation (SOKC). Per perform a Monte Carlo simulation for representative cases. Given the knowledge correlation SR QU-A3, the effect of the SOKC has been found to reference to ISLOCA frequency in the ASME Standard and the finding, Page 37 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion MU-D1-01 Finding QU-A1-01 Suggestion QU-A3-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text accordance with the does not specifically require perfonnance of a peer standard for PSA model upgrades.

requirements specified in review for upgrades.

Finding Resolved Section 6 of the ASME PRA Standard) for those aspects The Standard specifically calls for a peer review for of the PRA that have been PRA Upgrades, but the Configuration Control upgraded. Refer to Section Notebook does not specifically call for one following 2 of the ASME PRA an Upgrade.

Standard for the distinction Modify the Configuration Control Notebook to specify of a PRA Upgrade versus that peer reviews are required for PRA Upgrades.

PRA maintenance and update.

The PRA configuration The Configuration Control Notebook does not direct Section 3.3 of the configuration control notebook NB-PSA-CC [17] has control process shall include that updates or upgrades are compared with previous been revised to include requirements for the review of updates and evaluation of the impact of risk-infonned decisions and have used the PRA.

upgrades against previous applications and analyses.

changes on previously Review of previous RI applications is not called out in Finding Resolved implemented risk-infonned the Configuration Control Notebook.

decisions that have used the PRA AND that affect the Add requirement for reviewing the previous RI safe operation of the plant.

applications against the new PRA results to see if they impact the results of the previous work.

INTEGRATE the accident Figure 5-1 of the Quantification Report provides a To be implemented in a future notebook revision. This is a documentation sequences, system models, small flow chart on the process of integrating the issue only. No impact to the interim FPIE application for RI-ISI screening.

data, and HRA in the CAFTA models into the SAPHIRE code and the Suggestion Open quantification process for additional APls used to prepare the SAPHIRE model each initiating event group, for quantification (including the integration of CCF accounting for system trees, HRA rules, etc.) While this flow chart gives an dependencies, to arrive at upper level explanation of the process, a more accident sequence detailed flow chart would be useful in ensuring a frequencies.

consistent integration for personnel that do not perfonn this task frequently.

The integration process is fairly complex and involves multiple codes and tools. Missing any step in this process could impact quantification.

Develop a more detailed flow chart for those perfonning the quantification.

ESTIMATE the mean CDF The mean ISLOCA CDF frequency does not account A method of demonstrating the effect of the state of knowledge is to accounting for the state-of-for the state-of-knowledge correlation (SOKC). Per perfonn a Monte Carlo simulation for representative cases. Given the knowledge correlation SR QU-A3, the effect of the SOKC has been found to reference to ISLOCA fr~quency in the ASME Standard and the finding, Page 37 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions (Supporting Fmdingor C

1.200 Finding Description (summary discussion)

Disposition Requirement) between event probabilities be significant in cutsets contributing to ISLOCA two examples were selected using as input the failure rate and when significant [Note (1)].

frequency. Explicitly required in Note 1 of the SR.

distributions from the PSAR3 model, namely: ECCS injection line check Update the ISLOCA frequencies with SOKC.

valves FTRC and SDC MOVs FTRC. Each of these leads to an ISLOCA.

Based on these simulations, a correction factor was applied as a recovery event to the ISLOCA cut sets generated by CAFTA containing the following components and failure modes when generating results with point factors:

2 MOVs FTRC

- SOKC factor =3 (SDC suction) 2 check valves FTRC - SOKC factor = 4 (HPSI and LPSI injection lines) 3 check valves FTRC

- SOKC factor = 33 (HPSI injection lines)

The result is an increase in the initiating event frequency by a factor of 3.

The LPSI injection and SDC lines dominate the ISLOCA results. A factor of 3 (SDC) or 4 (LPSI) is not a significant deviation, particularly for applications where uncertainty analyses are performed as a part of the evaluation. Because incorporating the suggested rules file into the SAPH IRE model results in a negligible impact on overall core damage frequency ( within the uncertainty of the analysis), the event tree rules and basic events developed here to account for the SOKC will only be incorporated into the model for specific applications that examine ISLOCA events.

Finding Resolved QU-B2-01 Finding TRUNCATE accident Palisades used a truncation level of 1 E-09 for An analysis to determine the full power internal events model truncation sequences and associated quantification and conducted evaluation of limit will be documented in Section 6.6 of PSA notebook NB-PSA-QU after system models at a convergence of the results down to a truncation level issuance of the complete model report, including human error sufficiently low cutoff value of 1 E-1 2. The truncation should be set to 1 E-l 1 dependency, and full cutset review of all event trees.

that dependencies based on the Palisades definition of significant This approach will apply the ASME PRA standard HLR-QU-B which associated with significant accident sequences.

states, convergence can be considered sufficient when successive cutsets or accident reductions in truncation value of one decade result in decreasing changes sequences are not in CDF or LERF, and the final change is less than 5% which indicate that eliminated. NOTE:

a truncation of four orders of magnitude below the CDF is adequate for a Truncation should be high quality PRA.

carefully assessed in cases where cutsets are merged to A truncation study has been completed for the flooding PRA as create a solution (e.g.,

documented in EA-PSA-INTFLOOD-13-06 Rev. 0 [43].

where system level cutsets are merged to create sequence level cutsets).

Finding Open Page 38 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion QU-B2-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text between event probabilities be significant in cutsets contributing to ISLOCA two examples were selected using as input the failure rate and when significant [Note (1 )].

frequency. Explicitly required in Note 1 of the SA.

distributions from the PSAR3 model, namely: ECCS injection line check Update the ISLOCA frequencies with SOKC.

valves FTRC and SOC MOVs FTRC. Each of these leads to an ISLOCA.

Based on these simulations, a correction factor was applied as a recovery event to the ISLOCA cut sets generated by CAFT A containing the follOwing components and failure modes when generating results with point factors:

2 MOVs FTRC - SOKC factor = 3 (SOC suction) 2 check valves FTRC - SOKC factor = 4 (HPSI and LPSI injection lines) 3 check valves FTRC - SOKC factor = 33 (HPSI injection lines)

The result is an increase in the initiating event frequency by a factor of 3.

The LPSI injection and SOC lines dominate the ISLOCA results. A factor of 3 (SOC) or 4 (LPSI) is not a significant deviation, particularly for applications where uncertainty analyses are perfonned as a part of the evaluation. Because incorporating the suggested rules file into the SAPHIRE model results in a negligible impact on overall core damage frequency (within the uncertainty of the analysis), the event tree rules and basic events developed here to account for the SOKC will only be incorporated into the model for specific applications that examine ISLOCA events.

Finding Resolved TRUNCATE accident Palisades used a truncation level of 1 E-09 for An analysis to detennine the full power intemal events model truncation sequences and associated quantification and conducted evaluation of limit will be documented in Section 6.6 of PSA notebook NB-PSA-QU after system models at a convergence of the results down to a truncation level issuance of the complete model report, including human error sufficiently low cutoff value of 1 E-12. The truncation should be set to 1 E...,11 dependency, and full cutset review of all event trees.

that dependencies based on the Palisades definition of significant This approach will apply the ASME PRA standard HLR-QU-B which associated with significant accident sequences.

states, "convergence can be considered sufficient when successive cutsets or accident reductions in truncation value of one decade result in decreasing changes sequences are not in CDF or LERF, and the final change is less than 5% which indicate that eliminated. NOTE:

a truncation of four orders of magnitude below the CDF is adequate for a Truncation should be high quality PRA".

carefully assessed in cases A truncation study has been completed for the flooding PRA as where cutsets are merged to create a solution (e.g.,

documented in EA-PSA-INTFLOOD-13-06 Rev. 0 [43].

where system level cutsets are merged to create Finding Open sequence level cutsets).

Page 38 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement)

QU-C1-01 Finding IDENTIFY cutsets with Conditional HEPs were developed by Palisades for The complete detailed methodology for evaluating human error multiple HFEs that several HFEs and incorporated in the fault tree dependency is described in HRA Notebook NB-PSA-HR Volume 1 [25].

potentially impact significant models. Some accident sequences revealed HFE The dependency analysis for fire related HEPs are documented in section accident sequences/ cutsets combinations for which dependency between the 6.4 of the notebook.

by requantifying the PRA HFE5 has not been assessed and documented.

The general steps used in this analysis are as follows:

model with HEP values set

1. Run the base model with the post-initiator action failure event to values that are sufficiently probabilities set to 1.0.

high that the cutsets are not While the Palisades model has been quantified and truncated. The final cut sets for accident sequences have been identified,

2. Identify the multiple human action combinations that appear in the cut quantification of these post-the review and update of those sequences with sets.

initiator HFEs may be done respect to combinations of HFEs is not complete.

3. Identify the risk significant combinations assuming complete at the cutset level or saved dependence.

sequence level.

4. Perform a dependency analysis on the risk significant combinations Complete review and update of accident sequence and develop conditional probabilities for dependent actions.

cut sets relating to combinations of HFEs.

5. Incorporate the dependent combinations in the fault trees of the PSA.

To address the human action dependency issue with respect to CDF, Palisades developed a systematic approach that investigated a sufficient number of human actions to merit confidence that the impact of these dependencies have been thoroughly assessed and adequately represented in the PSA models. The approach is iterative and methodical.

This process has not been implemented for FPIE specific HEPs.

A human error dependency analysis was completed for the flooding PRA as documented in EA-PSA-INTFLOOD-13-06 Rev. 0 [43]. This finding is still open for the full power internal events PRA.

Finding Open for full power intemal events HEPs.

QU-D1-01 Finding REVIEW a sample of the The final model review has not been completed and The documentation of the final model is complete with exception of the significant accident documented. The final review of accident sequence final results cutset review for the remaining full power internal events sequences/cutsets sufficient results has not been completed and documented so initiators with human error dependency incorporated. Final, validation will to determine that the logic of that the reasonableness of the results can be verified, comport to the guidelines cited and Entergy procedures referenced in PSA the cutset or sequence is Palisades indicated that this review is required but not Notebook NB-PSA-CC [6].

correct.

complete. This finding is being written against all of the QU-D supporting requirements as well as some The updated flooding PRA is fully documented in EA-PSA-INTFLOOD-13-06 Rev. 0 [43].

QU-F requirements. Palisades needs to complete the formal review of accident sequence quantification results and make modifications as needed to address issues found in that review. The final results should Finding Open then be documented in the corresponding notebooks.

Page 39 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion QU-C1-01 Finding QU-D1-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text IDENTIFY cutsets with Conditional HEPs were developed by Palisades for The complete detailed methodology for evaluating human error multiple HFEs that several HFEs and incorporated in the fault tree dependency is described in HRA Notebook NB-PSA-HR Volume 1 [25].

potentially impact significant models. Some accident sequences revealed HFE The dependency analysis for fire related HEPs are documented in section I

accident sequencesl cutsets combinations for which dependency between the 6.4 of the notebook.

by requantifying the PRA HFEs has not been assessed and documented.

The general steps used in this analysis are as follows:

model with HEP values set

1. Run the base model with the post-initiator action failure event to values that are sufficiently probabilities set to 1.0.

high that the cutsets are not While the Palisades model has been quantified and truncated. The final cut sets for accident sequences have been identified,

2. Identify the multiple human action combinations that appear in the cut quantification of these post-the review and update of those sequences with sets.

initiator HFEs may be done respect to combinations of HFEs is not complete.

3. Identify the risk significant combinations assuming complete at the cutset level or saved dependence.

sequence level.

4. Perform a dependency analysiS on the risk Significant combinations Complete review and update of accident sequence and develop conditional probabilities for dependent actions.

cut sets relating to combinations of HFEs.

5. Incorporate the dependent combinations in the fault trees of the PSA.

To address the human action dependency issue with respect to CDF, Palisades developed a systematic approach that investigated a sufficient number of human actions to merit confidence that the impact of these dependencies have been thoroughly assessed and adequately represented in the PSA models. The approach is iterative and methodical.

This process has not been implemented for FPIE specific HEPs.

A human error dependency analysis was completed for the flooding PRA as documented in EA-PSA-INTFLOOD-13-06 Rev. 0 [43]. This finding is still open for the full power intemal events PRA.

Finding Open for full power internal events HEPs.

REVIEW a sample of the The final model review has not been completed and The documentation of the final model is complete with exception of the significant accident documented. The final review of accident sequence final results cutset review for the remaining full power intemal events sequences/cutsets sufficient results has not been completed and documented so initiators with human error dependency incorporated. Final, validation will to determine that the logic of that the reasonableness of the results can be verified.

comport to the guidelines cited and Entergy procedures referenced in PSA the cutset or sequence is Palisades indicated that this review is required but not Notebook NB-PSA-CC [6].

correct.

complete. This finding is being written against all of The updated flooding PRA is fully documented in the QU-D supporting requirements as well as some QU-F requirements. Palisades needs to complete the EA-PSA-INTFLOOD-13-06 Rev. 0 [43].

formal review of accident sequence quantification results and make modifications as needed to address Finding Open issues found in that review. The final results should then be documented in the corresponding notebooks.

Page 39 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

SPECIFY an appropriate mission time for the modeled accident sequences. For sequences in which stable plant conditions have been achieved, USE a minimum mission time of 24 hr.

Mission times for individual SSCs that function during the accident sequence may be less than 24 hr, as long as an appropriate set of SSCs and operator actions are modeled to support the full sequence mission time.

For example, if following a LOCA, low pressure injection is available for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, after which recirculation is required, the mission time for LPSI may be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and the mission time for recirculation may be 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />. For sequences in which stable plant conditions would not be achieved by 24 hr using the modeled plant equipment and human actions, PERFORM additional evaluation or modeling by using an appropriate technique.

Examples of appropriate techniques include: (a) assigning an appropriate plant damage state for the sequence; (b) extending the mission time, and adjusting the affected analyses, to the point at which conditions can be shown to reach Palisades uses 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as the default mission time for all sequences that end in a stable end state. This can be potentially overly conservative for some sequences such as LOOP sequences when power is not recovered by 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. A recovery factor considering the convolution of EDG FIR with offsite power was used but did not account for increased time for recovery as a function of the time that the EDG could run before failure Using 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for FTR in some sequences overestimates the importance of some events.

Potentially adjust the EDG FTR recovery factor to credit the increased time available for recovery of offsite power as a function of how long the EDG runs before failure.

Palisades has re-evaluated the data analysis and the time dependent models for the treatment of LOOP events. The modeling aspects include the time of LOOP recovery, the time of onsite power system recovery, EDG mission time, and the coping time between the time of an SBO event and the time when electric power must be recovered to prevent core damage. In addition to analyzing these interactions, the August 2003 northeast blackout event is evaluated in the data analysis as well.

The approach used to develop appropriate recovery factors in the analysis is as follows:

• A best estimate model is adopted based on the NUREG/CR-6890 curve fit for all types of LOOP events based on all 105 events.

• To account for uncertainty in the model and curve fitting and to account for the sparcity of data, an upper bound model and lower bound model were also devised. These upper and lower bound models are established by the following relationships:

• The upper bound, best estimate, and lower bound models are assumed to coincide at t=0

• At t=10 hours the Upper Bound Model is a factor of 2 higher than the best estimate model and the Lower Bound Model is a factor of 2 lower than the best estimate model in terms of the cumulative non exceedance probability of the time to recover offsite power.

• At t=20 hours the Upper Bound Model is a factor of 3 higher than the best estimate model and the Lower Bound Model is a factor of 3 lower than the best estimate model in terms of the cumulative non exceedance probability of the time to recover offsite power.

• The factor difference between the upper and lower bounds and the best estimate is permitted to grow in a log-linear fashion from 0 to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, from 10 to 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, and beyond 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. This means that the curve fit uncertainty factor is allowed to grow uniformly with time over the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

• In the uncertainty analysis, a discrete distribution over the three curves is used with 10% probability assigned to the Upper and Lower Bound Models, and 80% probability to the Best Estimate Model.

Suggestion Resolved SC-A5-01 Suggestion Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Requirement)

Suggestion Category II Text Page 40 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion SC-AS-01 Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i}

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text SPECIFY an appropriate Palisades uses 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as the default mission time Palisades has re-evaluated the data analysis and the time dependent mission time for the for all sequences that end in a stable end state. This models for the treatment of LOOP events. The modeling aspects include modeled accident can be potentially oveny conservative for some the time of LOOP recovery, the time of onsite power system recovery, sequences. For sequences sequences such as LOOP sequences when power is EDG mission time, and the coping time between the time of an SBO event in which stable plant not recovered by 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. A recovery factor and the time when electric power must be recovered to prevent core conditions have been considering the convolution of EDG FTR with offsite damage. In addition to analyzing these interactions, the August 2003 achieved, USE a minimum power was used but did not account for increased northeast blackout event is evaluated in the data analysis as well.

mission time of 24 hr.

time for recovery as a function of the time that the The approach used to develop appropriate recovery factors in the analysis Mission times for individual EDG could run before failure SSCS that function during is as follows:

the accident sequence may Using 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for FTR in some sequences

- A best estimate model is adopted based on the NUREG/CR-6890 curve be less than 24 hr, as long overestimates the importance of some events.

fit for all types of LOOP events based on all 105 events.

as an appropriate set of Potentially adjust the EDG FTR recovery factor to

- To account for uncertainty in the model and curve fitting and to account SSCs and operator actions credit the increased time available for recovery of are modeled to support the offsite power as a function of how long the EDG runs for the sparcity of data, an upper bound model and lower bound model full sequence mission time.

before failure.

were also devised. These upper and lower bound models are For example, if following a established by the following relationships:

LOCA, low pressure

-The upper bound, best estimate, and lower bound models are injection is available for 1 assumed to coincide at t=O hour, after which

-At t=1 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> the Upper Bound Model is a factor of 2 higher than recirculation is required, the mission time for LPSI may the best estimate model and the Lower Bound Model is a factor of 2 be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and the mission lower than the best estimate model in terms of the cumulative non-time for recirculation may be exceedance probability of the time to recover offsite power.

23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />. For sequences in

-At t=20 hours the Upper Bound Model is a factor of 3 higher than which stable plant conditions the best estimate model and the Lower Bound Model is a factor of 3 would not be achieved by 24 lower than the best estimate model in terms of the cumulative non-hr using the modeled plant exceedance probability of the time to recover offsite power.

equipment and human

-The factor difference between the upper and lower bounds and the actions, PERFORM additional evaluation or best estimate is permitted to grow in a log-linear fashion from 0 to modeling by using an 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, from 10 to 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, and beyond 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. This means appropriate technique.

that the curve fit uncertainty factor is allowed to grow uniformly with Examples of appropriate time over the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

techniques include: (a)

- In the uncertainty analysis, a discrete distribution over the three assigning an appropriate curves is used with 10010 probability assigned to the Upper and plant damage state for the Lower Bound Models, and 80% probability to the Best Estimate sequence; (b) extending the Model.

mission time, and adjusting Suggestion Resolved the affected analyses, to the point at which conditions can be shown to reach Page 40 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

Finding or ASME Reg. Guide 1.200 (Supporting Suggestion Category II Text Finding Description (summary discussion)

Disposition Requirement) acceptable values; or (c) modeling additional system recovery or operator actions for the sequence, in accordance with requirements stated in Systems Analysis (2-2.4) and Human Reliability (2-2.5) to demonstrate that a successful outcome is achieved.

SC-B5-01 Suggestion CHECK the reasonableness Although the success criteria appear to be reasonable Section 10.0 and Table 10.0-1 were inserted in notebook NB-PSA-ETSC and acceptability of the and consistent, there was no documented evidence

[1 1]. This section describes a comparison of the Palisades success results of the that they had been checked against genenc or other criteria to some comparable event tree headings developed for Waterford thermal/hydraulic, structural, plants. Palisades did provide some documentation on 3, which is a similar Combustion Engineering designed PWR. The review or other supporting how the success criteria were developed and how concludes that there are no significant outliers in the success criteria engineering bases used to they compared to Combustion Engineering Owners between the two plants that cannot be attributed to design differences.

support the success criteria.

Group guidance but there was no single, centralized Examples of methods to set of documentation to demonstrate how Palisades Suggestion Resolved achieve this include: (a) met the comparison requirement of the SR. Palisades comparison with results of needs to provide documentation of the comparison to the same analyses other generic or similar plants or provide a set of performed for similar plants, references to other documents that support this accounting for differences in requirement.

unique plant features (b) comparison with results of similar analyses performed with other plant-specific codes (c) check by other means appropriate to the particular analysis SC-C2-01 Suggestion DOCUMENT the processes LOCA break sizes are given in detail. However, the The primary technical basis reference is included in Attachment 2 of NB-used to develop overall PRA traceability of the references provided for where and PSA-lE Rev. 4. The thermal hydraulic basis was developed in PLP0247-success criteria and the how these break sizes were determined is difficult to 07-0004.01 R2, Palisades Nuclear Plant Thermal Hydraulic MAAP supporting engineering follow to the ultimate basis. Based on discussion with Calculations [14]. Additional description and technical basis is contained bases, including the inputs, the lead PRA Engineer, a reference is available for in EA-PSA-LOCA-IE-12-02, Initiating Event Frequencies for Loss of methods, and results. For these break sizes.

Coolant Accidents for the Palisades Nuclear Plant Probabilistic Risk example, this documentation Assessment [30]. References to these documents were added to Section typically includes: (a) the Documentation only.

5.0 of NB-PSA-ETSC [1 1].

definition of core damage Include a reference in the success criteria notebook Page 41 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion SC-BS-01 Suggestion SC-C2-01 Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text acceptable values; or (c) modeling additional system recovery or operator actions for the sequence, in accordance with requirements stated in Systems Analysis (2-2.4) and Human Reliability (2-2.5) to demonstrate that a successful outcome is achieved.

CHECK the reasonableness Although the success criteria appear to be reasonable Section 10.0 and Table 10.0-1 were inserted in notebook NB-PSA-ETSC and acceptability of the and consistent, there was no documented evidence

[11]. This section describes a comparison of the Palisades success results of the that they had been checked against generic or other criteria to some comparable event tree headings developed for Waterford thermallhydraulic, structural, plants. Palisades did provide some documentation on 3, which is a similar Combustion Engineering designed PWA. The review or other supporting how the success criteria were developed and how concludes that there are no significant outliers in the success criteria engineering bases used to they compared to Combustion Engineering Owners between the two plants that cannot be attributed to design differences.

support the success criteria.

Group guidance but there was no single, centralized Suggestion Resolved Examples of methods to set of documentation to demonstrate how Palisades achieve this include: (a) met the comparison requirement of the SA. Palisades comparison with results of needs to provide documentation of the comparison to the same analyses other generic or similar plants or provide a set of performed for similar plants, references to other documents that support this accounting for differences in requirement.

unique plant features (b) comparison with results of similar analyses performed with other plant-specifiC codes (c) check by other means appropriate to the particular analysis DOCUMENT the processes LOCA break sizes are given in detail. However, the The primary technical basis reference is included in Attachment 2 of NB-used to develop overall PRA traceability of the references provided for where and PSA-IE Rev. 4. The thermal hydraulic basis was developed in PLP0247-success criteria and the how these break sizes were determined is difficult to 07-0004.01 R2, "Palisades Nuclear Plant Thermal Hydraulic MAAP supporting engineering follow to the ultimate basis. Based on discussion with Calculations" [14]. Additional description and technical basis is contained bases, including the inputs, the lead PRA Engineer, a reference is available for in EA-PSA-LOCA-IE-12-02, "Initiating Event Frequencies for Loss of methods, and results. For these break sizes.

Coolant Accidents for the Palisades Nuclear Plant Probabilistic Risk example, this documentation Documentation only.

Assessment" [30]. References to these documents were added to Section typically includes: (a) the 5.0 of NB-PSA-ETSC [11].

definition of core damage Include a reference in the success criteria notebook Page 41 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Requirement)

Suggestion Category II Text Finding Description (summary discussion)

Disposition used in the PRA including that shows how the LOCA break sizes were Suggestion Resolved the bases for any selected determined.

parameter value used in the definition (e.g., peak cladding temperature or reactor vessel level) (b) calculations (generic and plant-specific) or other references used to establish success criteria, and identification of cases for which they are used (c) identification of computer codes or other methods used to establish plant-specific success criteria (d) a description of the limitations (e.g., potential conservatisms or limitations that could challenge the applicability of computer models in certain cases) of the calculations or codes (e) the uses of expert judgment within the PRA, and rationale for such uses (f) a summary of success criteria for the available mitigating systems and human actions for each accident initiating group modeled in the PRA (g) the basis for establishing the time available for human actions (h) descriptions of processes used to define success criteria for grouped initiating events or accident sequences SC-C3-01 Finding DOCUMENT the sources of Some Calculations associated with success criteria With regard to success criteria, the technical reference is documented in model uncertainty and are not in the Palisades formal document control the event tree and success criteria notebook: PLP0247-07-0004.01 RO, related assumptions (as system. In addition, the basis for the LOCA size Palisades Nuclear Plant Thermal Hydraulic MAAP calculations (R-1551).

Page 42 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion SC-C3-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text used in the PRA including that shows how the LOCA break sizes were Suggestion Resolved the bases for any selected determined.

parameter value used in the definition (e.g., peak cladding temperature or reactor vessel level) (b) calculations (generic and plant-specific) or other references used to establish success criteria, and identification of cases for which they are used (c) identification of computer codes or other methods used to establish plant-specific success criteria (d) a description of the limitations (e.g., potential conservatisms or limitations that could challenge the applicability of computer models in certain cases) of the calculations or codes (e) the uses of expert judgment within the PRA, and rationale for such uses (f) a summary of success criteria for the available mitigating systems and human actions for each accident initiating group modeled in the PRA (g) the basis for establishing the time available for human actions (h) descriptions of processes used to define success criteria for grouped initiating events or accident sequences DOCUMENT the sources of Some Calculations associated with success criteria With regard to success criteria, the technical reference is documented in model uncertainty and are not in the Palisades formal document control the event tree and success criteria notebook: PLP0247-07-0004.01 RO, related assumptions (as system. In addition, the basis for the LOCA size Palisades Nuclear Plant Thermal Hydraulic MAAP calculations (R-1551).

Page 42 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)Q)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Suggestion Category II Text Requirement) identified in QU-El and QU-ranges are not included. The issue regarding the Additional discussion and basis regarding LOCA size and frequency E2) associated with the basis for the LOCA size definitions is briefly repeated determination is contained in calculation EA-PSA-IE-00-0010, Revision development of success in suggestion F&0 SC-C2-01.

0,Calculation of Initiating Event Frequencies in Accordance with CEOG criteria.

Standards.

The formal document control system is predicated on an approved licensing action (e.g., Submittal of a EA-PSA-IE-00-0010 was included as a reference in notebook License Amendment Request for NFPA 805 or a NB-PSA-ETSC to improve the discussion regarding the basis for the power uprate). Therefore, some calculations are not determination of LOCA size ranges. These references were included in formally added to the system until the final project the overall set of documents provided to the peer review team on action is complete. This leaves some calculations 10/26/09.

used to support PRA success criteria out of the It is worth noting that during the Palisades PTS study [27], a separate system for some time and could result in lost or effort was underway at NRC to review and revise the LOCA frequencies modified documentation that does not comport with from NUREG/CR-5750 for use particularly in work associated with the PRA results. Technical bases for the size ranges 10CFR5O.46 but with applicability for other risk-informed applications such are not included in the success criteria definitions.

as the PTS project. There was a concem that the LOCA frequencies in NUREG/CR-5750 did not account for age-related factors important to deriving the frequencies and an expert elicitation effort at NRC was conducted to account for these adjustments.

Examining just the piping contribution it was concluded by the NRC Expert Elicitation committee that the Palisades plant specific initiating event frequencies were nearly the same as that developed in the elicitation effort. Therefore no change was made to the Palisades values during the conduct of the PTS analysis.

And the current Palisades small break LOCA frequency 2.26E-03/yr is approximately an order of magnitude greater than that reported in NUREG/CR-6928 mean value of 5.77E-04/yr.

In summary the Palisades LOCA frequencies are well documented and validated.

With respect to design processes, the site process for formal document control is being followed. There is not an elevated potential for lost or modified documentation that does not comport with the PRA results since the new PRA results are not formal results until the entire engineering change related to the submittal is complete.

The PRA staff is required to follow the plants design authority rules.

In conclusion, from a technical context and process assessment perspective this is not considered a finding.

Finding Resolved SY-A13-01 Finding INCLUDE those failures that Currently a flow diversion pathway is modeled for the Gates FLW-DIV-P54B&C-INJ, FLOW DIVERSION TRHOUGH P-54B can cause flow diversion Containment Spray pumps failing due to a diversion AND P-54C DURING INJECTION MODE, FLW-DIV-P54A&B-INJ FLOW pathways that result in through a failed other Containment Spray pump with a DIVERSION TRHOUGH P-54A AND P-54B DURING INJECTION failure to meet the system failed outboard check valve. Although this is a valid MODE, and FLW-DIV-P54A&C-INJ FLOW DIVERSION TRHOUGH Page 43 of 55 F&O#

Finding or (Supporting Suggestion Requirement)

SY-A13-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Category II Text Finding Description (summary discussion)

Disposition identified in aU-E1 and au-ranges are not included. The issue regarding the Additional discussion and basis regarding LOCA size and frequency E2) associated with the basis for the LOCA size definitions is briefly repeated determination is contained in calculation EA-PSA-IE-00-0010, Revision development of success in suggestion F&O SC-C2-01.

0, "Calculation of Initiating Event Frequencies in Accordance with CEOG criteria.

The formal document control system is predicated on Standards".

an approved licensing action (e.g., Submittal of a EA-PSA-IE-OO-OO10 was included as a reference in notebook License Amendment Request for NFPA 805 or a NB-PSA-ETSC to improve the discussion regarding the basis for the power uprate). Therefore, some calculations are not determination of LOCA size ranges. These references were included in formally added to the system until the final project the overall set of documents provided to the peer review team on action is complete. This leaves some calculations 10/26/09.

used to support PRA success criteria out of the It is worth noting that during the Palisades PTS study [27], a separate system for some time and could result in lost or modified documentation that does not comport with effort was underway at NRC to review and revise the LOCA frequencies the PRA results. Technical bases for the size ranges from NUREGlCR-5750 for use particularly in work associated with are not included in the success criteria definitions.

10CFR50.46 but with applicability for other risk-informed applications such as the PTS project. There was a concern that the LOCA frequencies in NUREG/CR-5750 did not account for age-related factors important to deriving the frequencies and an expert elicitation effort at NRC was conducted to account for these adjustments. Examining just the piping contribution it was concluded by the NRC Expert Elicitation committee that the Palisades plant specific initiating event frequencies were nearly the same as that developed in the elicitation effort. Therefore no change was made to the Palisades values during the conduct of the PTS analysis.

And the current Palisades small break LOCA frequency 2.26E-03/yr is approximately an order of magnitude greater than that reported in NUREG/CR-6928 mean value of 5.nE-04/yr. In summary the Palisades LOCA frequencies are well documented and validated.

With respect to design processes, the site process for formal document control is being followed. There is not an elevated potential for lost or modified documentation that does not comport with the PRA results since the new PRA results are not formal results until the entire engineering change related to the submittal is complete.

The PRA staff is required to follow the plants design authority rules.

In conclusion, from a technical context and process assessment perspective this is not considered a finding.

Finding Resolved INCLUDE those failures that Currently a flow diversion pathway is modeled for the Gates FLW-DIV-P54B&C-INJ, "FLOW DIVERSION TRHOUGH P-54B can cause flow diversion Containment Spray pumps failing due to a diversion AND P-54C DURING INJECTION MODE", FLW-DIV-P54A&B-INJ "FLOW pathways that result in through a failed other Containment Spray pump with a DIVERSION TRHOUGH P-54A AND P-54B DURING INJECTION failure to meet the system failed outboard check valve. Although this is a valid MODE", and FLW-DIV-P54A&C-INJ "FLOW DIVERSION TRHOUGH Page 43 of 55 I

I

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition (Supporting Suggestion Category II Text Requirement) success criteria, flow diversion pathway during the injection mode of P-54A AND P-54C DURING INJECTION MODE were added to the PRA operation, it is not a flow diversion pathway during model (See Attachment E Section 7.3.2).

recirculation since the diverted flow would be diverted to the suction of the HPI pump

- which is These gates are coupled with house event ESS-HSE-RAS-PRE which is set to true for modeling fault trees applicable only to pre-RAS (injection) where the outlet of a portion of the Containment Spray (CS) flow is supposed to go anyway. Because mode of operation. When true, the flow diversion results in no flow from the affected containment spray pump.

the HPI pumps flow rate is a function of the pressure in the containment, it does not matter which CS path Finding Resolved provide the flow the pump, the total flow to/through HPI will not be impacted by the pathway. Therefore, the total flow from the operating CS pump to the CS spargers will also not be impacted.

Current modeling results in unnecessary conservatism.

Include this flow diversion pathway only for injection modes of operation and remove from the recirculation mode of operation.

SY-A20-01 Finding INCLUDE events Palisades specifically models planned activities Coincident unavailability was re-evaluated and updated in Section 9.1 of representing the resulting in coincident unavailability of equipment in the data analysis notebook, NB-PSA-DA [5]. From that evaluation:

simultaneous unavailability multiple trains of different systems that belong to To evaluate coincident unavailability, all the unavailability data was of redundant equipment similar divisions (such as train A of AFW and train A when this is a result of of HPI) but does not include events that might occur compiled, and coincident events were marked. Coincident unavailability was considered for each train (i.e., 2 or more train A components OOS at planned activity (see DA-associated with coincident unavailability of multiple C14).

trains of different systems that belong to opposite the same time), and for both trains (i.e., 1 or more Train A components OOS at the same time as 1 or more Train B components). In addition to division (such as train A of AFW and train B of HPI).

reviewing the maintenance rule unavailability data for coincident unavailability, the risk management work week reviews from the LAN were also downloaded and reviewed.

Potential unavailability between systems involving opposite divisions due to planned activities is not The following identifies the equipment associated with each train:

included in the model and may result in non conservative results.

• Train A equipment: C-2A & C-2C, C-6B, ED-15 & ED-17, K-6A, P-52C, P-54B & P-54C, P-55C, P-56A, P-66B, P-67B, P-7B, P-8A & P-8B, and PRy-i 042.

Include events in the model that address coincident

• Train B equipment: C-2B, C-6A, ED-16 & ED-18, K-6B, P-52B, P-54A, unavailabilities associated with train A of one system P-55A & P-55B, P-56B, P-66A, P-67A, P-7A & 7C, P-SC and PRy-i 043.

with train B of another, redundant systems due to Plant experience showed that in most cases only one piece of equipment Page 44 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion SY-A20-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text success criteria.

flow diversion pathway during the injection mode of P-54A AND P-54C DURING INJECTION MODE" were added to the PRA operation, it is not a flow diversion pathway during model (See Attachment E Section 7.3.2).

recirculation since the "diverted" flow would be These gates are coupled with house event ESS-HSE-RAS-PRE which is diverted to the suction of the HPI pump - which is where the outlet of a portion of the Containment set to true for modeling fault trees applicable only to pre-RAS (injection)

Spray (CS) flow is supposed to go anyway. Because mode of operation. When true, the flow diversion results in no flow from the HPI pumps flow rate is a function of the pressure the affected containment spray pump.

in the containment, it does not matter which CS path Finding Resolved provide the flow the pump, the total flow tolthrough HPI will not be impacted by the pathway. Therefore, the total flow from the operating CS pump to the CS spargers will also not be impacted.

Current modeling results in unnecessary conservatism.

Include this flow diversion pathway only for injection modes of operation and remove from the recirculation mode of operation.

INCLUDE events Palisades specifically models planned activities Coincident unavailability was re-evaluated and updated in Section 9.1 of representing the resulting in coincident unavailability of equipment in the data analysis notebook, NB-PSA-DA [5]. From that evaluation:

simultaneous unavailability multiple trains of different systems that belong to "To evaluate coincident unavailability, all the unavailability data was of redundant equipment similar divisions (such as train A of AFW and train A when this is a result of of HPI) but does not include events that might occur compiled, and coincident events were marked. Coincident unavailability planned activity (see DA-associated with coincident unavailability of multiple was considered for each train (i.e., 2 or more train A components OOS at C14).

trains of different systems that belong to opposite the same time), and for both trains (i.e., 1 or more Train A components division (such as train A of AFW and train B of HPI).

OOS at the same time as 1 or more Train B components). In addition to reviewing the maintenance rule unavailability data for coincident unavailability, the risk management work week reviews from the LAN were Potential unavailability between systems involving also downloaded and reviewed.

opposite divisions due to planned activities is not The follOwing identifies the equipment associated with each train:

included in the model and may result in non-

• Train A equipment: C-2A & C-2C, C-SB, ED-15 & ED-17, K-6A, P-52C, conservative results.

P-54B & P-54C, P-55C, P-5SA, P-66B, P-S7B, P-7B, P-8A & P-8B, and PRV-1042.

Include events in the model that address coincident

• Train B equipment: C-2B, C-SA, ED-16 & ED-18, K-6B, P-52B, P-54A, unavailabilities associated with train A of one system P-55A & P-55B, P-56B, P-66A, P-67A, P-7A & 7C, P-8C and PRV-1043.

with train B of another, redundant systems due to Plant experience showed that in most cases only one piece of equipment Page 44 of 55 I

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Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition (Supporting Suggestion Category II Text Requirement) planned activities (if the experience shows any exist).

from a train is removed from service at a time. A review of the three plus years of unavailability data showed that there was limited, repetitive coincident unavailability; most cases involved only two components, and occurred only once in the three year data window.

There were, however, a few cases in which plant experience showed that two components were recurrently removed from service at the same time.

In these cases, coincident unavailability was modeled; the following identifies the combinations of equipment for coincident unavailability:

1.

P-54B and P-66B; 2.

P-54B and P-67B; 3.

P-54C and P-67B; 4.

P-8A and P-8B; 5.

P-54A and P-66A; 6.

P-54A and P-67A; and 7.

P-56A and P-56B.

Basic events were developed for items 1-7 above and documented in 2, Table 12.1 of Reference [5].

Coincident unavailability included only the time that both components were simultaneously unavailable.

If one component was unavailable for an extra hour, the hour was used in the individual unavailability. Once coincident unavailabilitys were calculated, the times were subtracted from the individual unavailabilitys to avoid double counting.

Finding Resolved SY-B3-01 Finding ESTABLISH common cause Common cause failures as a whole are modeled A full evaluation of this finding is presented in Attachment 1 of Reference failure groups by using a correctly and consistently. However, the modeling of

[9].

logical, systematic process the HPI, LPI, and common line check valves is Examination of cut sets that include CCF of in-series components reveals that considers similarity in producing non-minimal and potentially non-valid that there are no non-minimal cut sets.

(a) service conditions (b) cutsets.

environment (c) design or Because of the safety significance of the LPI and HPI Treating in-series HPSI and LPSI valves as independent (incorporating manufacturer (d) the CCF portion of the valve failure in the failure probability for each maintenance JUSTIFY the systems, the non-minimal and non-valid cutsets are valve), as appears to be suggested by this finding, turns out to be the overestimating the risk associated with those failures.

basis for selecting common more conservative approach.

cause component groups.

Review the common cause modeling of components The Palisades approach produces realistic and valid results.

Candidates for common in the PRA model, especially of the valves in series cause failures include, for and revise the model as appropriate. Alternatively, The modeling of common cause failures, as applied in the Palisades PRA, Page 45 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion SY-B3-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text planned activities (if the experience shows any exist).

from a train is removed from service at a time. A review of the three plus years of unavailability data showed that there was limited, repetitive coincident unavailability; most cases involved only two components, and occurred only once in the three year data window.

There were, however, a few cases in which plant experience showed that two components were recurrently removed from service at the same time.

In these cases, coincident unavailability was modeled; the following identifies the combinations of equipment for coincident unavailability:

1.

P-54B and P-66B;

2.

P-54B and P-67B;

3.

P-54C and P-67B;

4.

P-8A and P-8B;

5.

P-54A and P-66A;

6.

P-54A and P-67 A; and

7.

P-56A and P-56B.

Basic events were developed for items 1-7 above and documented in 2, Table 12.1 of Reference [5].

Coincident unavailability included only the time that both components were simultaneously unavailable. If one component was unavailable for an extra hour, the hour was used in the individual unavailability. Once coincident unavailability's were calculated, the times were subtracted from the individual unavailability's to avoid double counting."

Finding Resolved ESTABLISH common cause Common cause failures as a whole are modeled A full evaluation of this finding is presented in Attachment 1 of Reference failure groups by using a correctly and consistently. However, the modeling of

[9].

logical, systematic process the HPI, LPI, and common line check valves is Examination of cut sets that include CCF of in-series components reveals that considers similarity in producing non-minimal and potentially non-valid that there are no non-minimal cut sets.

(a) service conditions (b) cutsets.

environment (c) design or Because of the safety significance of the LPI and HPI Treating in-series HPSI and LPSI valves as independent (incorporating manufacturer (d) systems, the non-minimal and non-valid cutsets are the CCF portion of the valve failure in the failure probability for each maintenance JUSTIFY the overestimating the risk associated with those failures.

valve), as appears to be suggested by this finding, turns out to be the basis for selecting common more conservative approach.

cause component groups.

Review the common cause modeling of components The Palisades approach produces realistic and valid results.

Candidates for common in the PRA model, especially of the valves in series cause failures include, for and revise the model as appropriate. Alternatively, The modeling of common cause failures, as applied in the Palisades PRA, Page 45 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

F d ASME R G *d 1 200 (Supporting Suggestkn categorre e

Finding Description (summary discussion)

Disposition Requirement) example: (a) motor-operated non-valid combinations can be added to the mutually is based on, and consistent with, the Multiple Greek Letter approach. This valves (b) pumps (c) safety-exclusive file to remove the non-minimal and non-approach produces valid cut sets, even if those cut sets may indicate that relief valves (d) air-operated valid cutsets.

more components have failed than necessary.

valves (e) solenoid-operated valves (f) check valves (ri)

The approximation suggested by this finding is, in fact, a more diesel generators (h) conservative approach which can overestimate risk.

If the beta factor is batteries (i) inverters and small, then this overestimation is not significant.

battery charger (j) circuit The approximation used in the Palisades PRA, namely, using the total breakers failure rate to represent the independent failure rate without correcting by the factor of (1-beta), also does not introduce significant conservatism in the results.

Therefore, the concerns expressed by this finding do not appear to be correct, and modeling or quantification changes are not considered necessary.

Finding Resolved SY-B4-01 Suggestion INCORPORATE common Because the CCF modeling approach for CCGs An evaluation of this peer review team suggestion was performed in cause failures into the greater than 5 is bounding, it is recommended that the of EA-PSA-RG1.200F&O-10-01 Rev. 2. In summary, the system model consistent impact of this conservatism be investigated in the evaluation concludes the following:

with the common cause sensitivity analysis.

model used for data

• The global CCF factor chosen for this evaluation for 8 components anal sis tailing due to common cause has a value that is higher (by a factor of y

more than 50) than the factor that is calculated using the explicit multiple Greek letter (MGL) approach.

• Atthe single component level, both the global and explicit approaches produce the same result. In other words, the bounding value of the global CCF factor is representative of the excluded combinations of components, and vice versa. This is the expected result.

• At the system level, the quantification of a system fault tree that incorporates the bounding global CCF event in addition to all other random and common cause failures that contribute to system failure suggests that the use of the global CCF factor is bounding, but small (on the order of several percent).

• For a global CCF to have a significant effect on the overall results of the PRA, it likely would need to have an effect on multiple redundant systems. Such global CCF events may exist in the Palisades PRA (e.g., station power transformers, sequencers) and should be examined for potential further refinement of the CCF when they impact the results of an application significantly.

Page 46 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion SY-84-01 Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text example: (a) motor-operated non-valid combinations can be added to the mutually is based on, and consistent with, the Multiple Greek Letter approach. This valves (b) pumps (c) safety-exclusive file to remove the non-minimal and non-approach produces valid cut sets, even if those cut sets may indicate that relief valves (d) air-operated valid cutsets.

more components have failed than necessary.

valves (e) solenoid-operated The approximation suggested by this finding is, in fact, a more valves (f) check valves (g) conservative approach which can overestimate risk. If the beta factor is diesel generators (h) small, then this overestimation is not significant.

batteries (i) inverters and battery charger 0) circuit The approximation used in the Palisades PRA, namely, using the "total" breakers failure rate to represent the "independenr failure rate without correcting by the factor of (1-beta), also does not introduce significant conservatism in the results.

Therefore, the concems expressed by this finding do not appear to be correct, and modeling or quantification changes are not considered necessary.

Finding Resolved INCORPORATE common Because the CCF modeling approach for CCGs An evaluation of this peer review team suggestion was perfonned in cause failures into the greater than 5 is bounding, it is recommended that the of EA-PSA-RG1.200F&O-10-01 Rev. 2. In summary, the system model consistent impact of this conservatism be investigated in the evaluation concludes the following:

with the common cause sensitivity analysis.

-The global CCF factor chosen for this evaluation for 8 components model used for data analysis.

failing due to common cause has a value that is higher (by a factor of more than 50) than the factor that is calculated using the explicit multiple Greek letter (MGL) approach.

-At the single component level, both the global and explicit approaches produce the same result. In other words, the bounding value of the global CCF factor is representative of the excluded combinations of components, and vice versa. This is the expected result.

-At the system level, the quantification of a "system" fault tree that incorporates the bounding global CCF event in addition to all other random and common cause failures that contribute to system failure suggests that the use of the global CCF factor is bounding, but small (on the order of several percent).

-For a global CCF to have a significant effect on the overall results of the PRA, it likely would need to have an effect on multiple redundant systems. Such global CCF events may exist in the Palisades PRA (e.g., station power transfonners, sequencers) and should be examined for potential further refinement of the CCF when they impact the results of an application significantly.

Page 46 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

F d ASME A G *d 1 200 (Supporting SuggesUon Category H Text e

Finding Description (summary discussion)

Disposition Requirement)

The impact of the bounding approach is to produce results that are conservative, with the amount of conservatism being a function of the component group and its MGL factors.

Suggestion Resolved SY-B5-01 Finding ACCOUNT explicitly for the There is an apparent error in the EDG failure to run This is not considered a finding. SWS start failures are captured under modeled systems logic: this logic does not account for the SWS pump the diesel failure to start gates. Start, load/run and run failures are all dependency on support failures to start. When the PRA Group was shown the captured under OR gates so the logic is equivalent.

systems or interfacing apparent error, they admitted that it was an error and systems in the modeling that they had also identified it in their Self-The PRA model Release 2b cutsets properly account for diesel run and process. This may be Assessment. The model was corrected while the service water pump start failures.

accomplished in one of the Review Team was on-site, but a review of the This issue was noted under supporting requirement QU-D5 in the Reg.

following ways: (a) for the affected cutsets still has a cutset with a diesel Guide 1.200 Self-Assessment (NB-PSA-SA Rev 0) for model Release 2a.

fault tree linking approach generator run failure in the same cutset as the SWS It was subsequently corrected in Release 2b delivered on 10/26/09 and by modeling the pump failure to start. Given failure of the SWS pump again noted in the updated Self-Assessment [18].

dependencies as a link to an to start, the diesel generator fail to run should be 1.0.

appropriate event or gate in Finding Resolved the support system fault tree; (b) for the linked event SWS pump failures to start are valid contributors to tree approach, by using EDG failure. The model should account for these event tree logic rules, or contributors and the diesel generator failures need to calculating a probability for be adjusted to account for the availability of SWS each split fraction conditional on the scenario definition.

These specific failures should be incorporated into the fault tree model. And, given the similarity of this finding with Finding SY-B5-02, it is recommended that a systematic review of other potentially risk important dependencies be performed.

SY-B5-02 Finding ACCOUNT explicitly for the Potentially risk-significant manual valves were Per supporting requirement SY-A15 [1]: A component may be excluded modeled systems excluded from the model without explanation. Their from the system model if the total failure probability of the component dependency on support exclusion should be based on SR SY-A15 screening failure modes resulting in the same effect on system operation is at least systems or interfacing criteria. For example, manual valves in the two orders of magnitude lower than the highest failure probability of the systems in the modeling Containment Spray system flow paths were not other components in the same system train that results in the same effect process. This may be modeled.

on system operation.

accomplished in one of the following ways: (a) for the fault tree linking approach It was noted that some of these manual valves are The valves described in finding SY-B5-02 in the containment spray system by modeling the actually depicted on the simplified system drawings, are normally locked open manual valves. The Palisades PRA has dependencies as a link to an but they are not labeled. To avoid confusion, it is assumed that random failure or plugging of locked open manual valves Page 47 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion SY-B5-01 Finding SY-B5-02 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text The impact of the bounding approach is to produce results that are conservative, with the amount of conservatism being a function of the component group and its MGL factors.

Suggestion Resolved ACCOUNT explicitly for the There is an apparent error in the EDG failure to run This is not considered a finding. SWS start failures are captured under modeled system's logic: this logic does not account for the SWS pump the diesel failure to start gates. Start, load/run and run failures are all dependency on support failures to start. When the PRA Group was shown the captured under 'OR' gates so the logic is equivalent.

systems or interfacing apparent error, they admitted that it was an error and The PRA model Release 2b cutsets properly account for diesel run and systems in the modeling that they had also identified it in their Self-process. This may be Assessment. The model was corrected while the service water pump start failures.

accomplished in one of the Review Team was on-site, but a review of the This issue was noted under supporting requirement aU-D5 in the Reg.

following ways: (a) for the affected cutsets still has a cutset with a diesel Guide 1.200 Self-Assessment (NB-PSA-SA Rev 0) for model Release 2a.

fault tree linking approach generator run failure in the same cutset as the SWS It was subsequently corrected in Release 2b delivered on 10/26/09 and by modeling the pump failure to start. Given failure of the SWS pump again noted in the updated Self-Assessment [18].

dependencies as a link to an to start, the diesel generator fail to run should be 1.0.

Finding Resolved appropriate event or gate in the support system fault tree; (b) for the linked event SWS pump failures to start are valid contributors to tree approach, by using EDG failure. The model should account for these event tree logic rules, or contributors and the diesel generator failures need to calculating a probability for be adjusted to account for the availability of SWS each split fraction conditional on the scenario definition.

These specific failures should be incorporated into the fault tree model. And, given the similarity of this finding with Finding SY -B5-02, it is recommended that a systematic review of other potentially risk important dependencies be performed.

ACCOUNT explicitly for the Potentially risk-significant manual valves were Per supporting requirement SY -A 15 [1]: A component may be excluded modeled system's excluded from the model without explanation. Their from the system model if the total failure probability of the component dependency on support exclusion should be based on SR SY -A 15 screening failure modes resulting in the same effect on system operation is at least systems or interfacing criteria. For example, manual valves in the two orders of magnitude lower than the highest failure probability of the systems in the modeling Containment Spray system flow paths were not other components in the same system train that results in the same effect process. This may be modeled.

on system operation.

accomplished in one of the following ways: (a) for the fault tree linking approach It was noted that some of these manual valves are The valves described in finding SY -B5-02 in the containment spray system by modeling the actually depicted on the simplified system drawings, are normally locked open manual valves. The Palisades PRA has dependencies as a link to an but they are not labeled. To avoid confusion, it is assumed that random failure or plugging of locked open manual valves Page 47 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions orting 1.200 Finding Description (summary discussion)

Disposition Requirement) appropriate event or gate in suggested that all components in these drawings be are not a significant contribution to system or component failure, however, the support system fault labeled. Note: site practice is to include all mechanical this assumption was not explicitly documented. Assumption number tree; (b) for the linked event components on the simplified PRA schematics and to A-0047 was developed and added to the PRA assumptions database, tree approach, by using label only those components specifically included, success criteria notebook [1 1], and appropriate system notebooks.

event tree logic rules or This provides a quick indication of what components calculating a probability for are physically present but not explicitly modeled.

The assumptions states:

each split fraction NUREG CR-6928 (January 2007) Table 5-1, provides data for manual conditional on the scenario valve failure to open and failure to close. Failure to remain open is not definition.

Excluded manual valves may be risk significant.

evaluated and no data is provided for this failure mode. Plugging has a mean failure probability of E-09. A valve locked in position, is very unlikely to be susceptible to environmental effects such as vibration that could Provide explanation for the excluded valves based on result in valve closure. In addition, locked valves are strictly controlled by SY-Al 5 or include them in the model.

keys issued from the control room and systems with locked open valves are either normally in operation or are frequently tested to meet technical specification requirements. A mispositioned or plugged valve would likely be detected as part of plant operator rounds, testing, check lists, or data collection and would be promptly re-positioned or repaired as necessary.

Based on the generic failure data for plugging, testing, and strict controls of these valves, the probability of valve failure is very small and would have a negligible impact on system failure rate.

This assumption is not applicable to pre-initiator human error events where a valve is repositioned for testing or maintenance.

With respect to manual valve misalignment, a scoping analysis was performed as documented in the Palisades HRA notebook NB-PSA-HR Volume 2, Palisades Pre-initiator Human Error Evaluation [8]. The results of this scoping demonstrated that a number of manual valves are susceptible to mispositioning with a non-significant failure probability. The basic events developed and scoping methodology are presented in that document.

A sampling of locked open and non-locking manual valves were evaluated from the auxiliary feedwater, shutdown cooling, and atmospheric dump valve fault trees to provide validation that assumption A-0047 is applied consistently to all system fault trees. No discrepancies were found.

Finding Resolved SY-B1 1 -01 Finding MODEL the ability of the The current model for the supplemental diesel model, A new fault tree was created, PNOSGPWR NO SAFEGUARDS POWER available inventories of air, however, is not completely correct as CB 152-203 TO SAFEGUARDS BUS that models failure of buses 1 C, 1 D, or 1 E and power, and cooling to should be fails to remain closed instead of Fails to failure of their respective breakers that tie them to the safeguards bus to support the mission time.

remain open, and failure of the A14 safeguards bus open.

needs_to_be_added_to_the_model_as_a_reason_the Page 48 of 55 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions r:u~:Orting

~inding ~r

~S~E Re~*TG~de 1.200 Rnding Description (summary discussion)

Disposition Requirement) ugges on a egory e

appropriate event or gate in suggested that all components in these drawings be are not a significant contribution to system or component failure, however, the support system fault labeled. Note: site practice is to include all mechanical this assumption was not explicitly documented. Assumption number tree; (b) for the linked event components on the simplified PRA schematics and to A-0047 was developed and added to the PRA assumptions database, tree approach, by using label only those components specifically included.

success criteria notebook [11], and appropriate system notebooks.

event tree logic rules, or This provides a quick indication of what components calculating a probability for are physically present but not explicitly modeled.

The assumptions states.

each split fraction "NUREG CR-6928 (January 2007) Table 5-1, provides data for manual conditional on the scenario valve failure to open and failure to close. Failure to remain open is not definition.

Excluded manual valves may be risk Significant.

evaluated and no data is provided for this failure mode. Plugging has a I

mean failure probability of E-09. A valve locked in position, is very unlikely!

to be susceptible to environmental effects such as vibration that could Provide explanation for the excluded valves based on result in valve closure. In addition, locked valves are strictly controlled by SY -A 15 or include them in the model.

keys issued from the control room and systems with locked open valves I

are either nonnally in operation or are frequently tested to meet technical I

specification requirements. A mispositioned or plugged valve would likely be detected as part of plant operator rounds, testing. check lists. or data collection and would be promptly re-positioned or repaired as necessary.

Based on the generic failure data for plugging. testing. and strict controls of these valves. the probability of valve failure is very small and would have a negligible impact on system failure rate.

This assumption is not applicable to pre-initiator human error events where a valve is repositioned for testing or maintenance."

With respect to manual valve misalignment. a scoping analysis was perfonned as documented in the Palisades HRA notebook NB-PSA-HR Volume 2. "Palisades Pre-initiator Human Error Evaluation" [8]. The results of this scoping demonstrated that a number of manual valves are susceptible to mispositioning with a non-significant failure probability. The basic events developed and scoping methodology are presented in that document.

A sampling of locked open and non-locking manual valves were evaluated from the auxiliary feedwater. shutdown cooling. and atmospheric dump valve fault trees to provide validation that assumption A-0047 is applied consistently to all system fault trees. No discrepancies were found.

Finding Resolved SY-B11-01 Rnding MODEL the ability of the The current model for the supplemental diesel model.

A new fault tree was created. PNOSGPWR "NO SAFEGUARDS POWER available inventories of air, however, is not completely correct as CB 152-203 TO SAFEGUARDS BUS" that models failure of buses 1 C. 1 D. or 1 E and power. and cooling to should be -fails to remain closed-instead of "Fails to failure of their respective breakers that tie them to the safeguards bus to support the mission time.

remain open-. and failure of the A14 safeguards bus open.

needs to be added to the model as a reason the ___ _ ~

Page 48 of 55

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Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions (Supporting Finding or 1.200 Finding Description (summary discussion)

Disposition Requirement)

Supplemental DG fails to provide power to the 1 D This fault tree was placed under gates:

Safeguards bus.

• PNOSGPWR1 D, NO SAFEGUARDS POWER TO BUS 1 D PNOSGPWR1C, NO POWER FROM SAFEGUARDS BUS TO BUS The 152-203 CB is modeled under another portion of 1C IogkD for power to the 1D Safeguards bus. Using

• PNOSGPWR1 E, NO SAFEGUARDS POWER TO BUS 1 E failure results in the impact for the failure of the CB Additional breaker failure logic was also added to model failures not being adequately captured in the model.

associated with the non-safety-related (NSR) diesel generator. This logic considers that the emergency diesel generator (EDG 1-2) and safety related bus supply breakers (1 C and 1 D) must open prior to starting the Revise the modeling to correct the CB failure mode NSR diesel generator and subsequently re-closing to supply the modeled, and add failure of the 1A bus itself to the appropriate bus.

model.

Under gate DG-NSR-START1D-03, CIRCUIT BREAKER FAILURES added basic events:

• ACP-C2MA-152-213, CIRCUIT BREAKER 152-213 FAILS TO OPEN

• ACP-C2MA-152-203, CIRCUIT BREAKER 152-203 FAILS TO OPEN

• ACP-C2MB-152-203, CIRCUIT BREAKER 152-203 FAILS TO CLOSE Under gate DG-NSR-RUN1D-03, CIRCUIT BREAKER FAILURES added basic events:

• ACP-C2MC-152-203, CIRCUIT BREAKER 152-203 FAILS TO REMAIN CLOSED Under gate DG-NSR-START1C-03, CIRCUIT BREAKER FAILURES added basic event:

• ACP-C2MB-152-403, NSR EDG OUTPUT BREAKER 152-403 FAILS TO CLOSE These logic changes capture all of the appropriate breaker failure modes related to the non-safety related emergency diesel generator and the safeguards bus.

Finding Resolved SY-B12-01 Finding DO NOT USE Palisades did not model HVAC for the control room or The basis for excluding control room HVAC from the full power internal proceduralized recovery the cable spreading room based on operator actions events model was strengthened to include other aspects in addition to actions as the sole basis for to implement alternate cooling strategies such as operator actions. This evaluation was fully documented in NB-PSA-ETSC eliminating a support system opening doors or using a proposed portable exhaust

[1 1]. The conclusion summary states:

from the model; however, fans. (See pages 17 and 24 of attachment 8 to NB-Page 49 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion SY-B12-01 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text Supplemental DG fails to provide power to the 1 D This fault tree was placed under gates:

Safeguards bus.

• PNOSGPWR1 D, NO SAFEGUARDS POWER TO BUS 1 D

• PNOSGPWR1C, NO POWER FROM SAFEGUARDS BUS TO BUS The 152-203 CB is modeled under another portion of 1C the logic for power to the 1 D Safeguards bus. USing

• PNOSGPWR1E, NO SAFEGUARDS POWER TO BUS 1E the incorrect failure modelbasic event for the CB failure results in the impact for the failure of the CB Additional breaker failure logic was also added to model failures not being adequat~ly captured in the model.

associated with the non-safety-related (NSR) diesel generator. This logic considers that the emergency diesel generator (EDG 1-2) and safety related bus supply breakers (1 C and 1 D) must open prior to starting the Revise the modeling to correct the CB failure mode NSR diesel generator and subsequently re-closing to supply the modeled, and add failure of the 1 A bus itself to the appropriate bus.

model.

Under gate DG-NSR-START1 D-03, "CIRCUIT BREAKER FAILURES" added basic events:

• ACP-C2MA-152-213, CIRCUIT BREAKER 152-213 FAILS TO OPEN

• ACP-C2MA-152-203, CIRCUIT BREAKER 152-203 FAILS TO OPEN

• ACP-C2MB-152-203, CIRCUIT BREAKER 152-203 FAILS TO CLOSE Under gate DG-NSR-RUN1 D-03, "CIRCUIT BREAKER FAILURES" added basic events:

• ACP-C2MC-152-203, CIRCUIT BREAKER 152-203 FAILS TO REMAIN CLOSED Under gate DG-NSR-START1C-03, "CIRCUIT BREAKER FAILURES" added basic event:

• ACP-C2MB-152-403, NSR EDG OUTPUT BREAKER 152-403 FAILS TO CLOSE These logic changes capture all of the appropriate breaker failure modes related to the non-safety related emergency diesel generator and the safeguards bus.

Finding Resolved DO NOT USE Palisades did not model HVAC for the control room or The basis for excluding control room HVAC from the full power intemal proceduralized recovery the cable spreading room based on operator actions events model was strengthened to include other aspects in addition to actions as the sole basis for to implement altemate cooling strategies such as operator actions. This evaluation was fully documented in NB-PSA-ETSC eliminating a support system opening doors or using a proposed portable exhaust

[11]. The conclusion summary states:

from the model; however, fans. (See pages 17 and 24 of attachment 8 to NB-Page 49 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions INCLUDE these recovery actions in the model quantification. For example, it is not acceptable to not model a system such as HVAC or CCW on the basis that there are procedures for dealing with losses of these systems.

PSA-ETSC rOl). However, the operator actions to implement the alternate actions were not included in the models. There was never an intent to model the operator actions given that past analyses has shown that both rooms can survive a loss of HVAC. It is recognized that the analyses requires updating and that the documentation requires updating.

Palisades should either provide additional justification for not modeling the HVAC systems for the cable spreading room and control room, or model the operator actions to implement alternate cooling strategies or model HVAC for these two rooms.

Control room cooling in the Palisades internal events PRA is not considered an issue based on the following:

• the high design temperature limits of the major control room components,

• the general conservative modeling assumptions employed throughout the EA-APR-95-023,R1 analysis,

• the philosophy of the operators with respect to remaining in the control room during such an event,

• the TMMs are not credited in the EOPs,

• the relative un-importance of HVAC failure on a variety of plant PRA studies.

Therefore it is considered unnecessary to model either loss of HVAC as an initiator or as a support system for the internal events model.

An analysis of the cable spreading room heat-up following a loss of ventilation was developed using the GOTHIC software code [12]. The analysis approach was to establish the rooms heat load based on Systematic Evaluation Program (SEP) Topic lX-5 (Phase II cable spreading room loss of HVAC testing) data by modeling the test boundary conditions in detail and iterating on room heat generation until the test results were mimicked by the model. With the room heat load established, the model boundary conditions were changed to establish a conservative scenario with no room ventilation. The room temperature profile demonstrated that at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> the peak temperature would reach 122°F (50°C).

CALC-455-001-DC2 [13] was then performed to evaluate all cable spreading room equipment modeled in the PRA under these conditions.

The analysis conservatively assumed that the room was at the peak calculated temperature of 122°F (50°C) for the entire 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> duration of the transient. An evaluation of equipment qualification reports, and vendor data, was then performed which concluded that reasonable assurance of operability is assured for all equipment at an elevated ambient of 122°F for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Based on the conclusions of these analyses, ventilation to the cable spreading area is not explicitly modeled as failure to re-establish ventilation does not result in equipment failure prior to the PRA 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time. of NB-PSA-ETSC [11] has been updated to reflect these F&O #

(Supporting Finding or ASME Beg. Guide 1.200 Requirement)

Suggestion Category II Text Finding Description (summary discussion)

Disposition Page 50 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text INCLUDE these recovery PSA-ETSC r01). However, the operator actions to Control room cooling in the Palisades internal events PRA is not actions in the model implement the altemate actions were not included in considered an issue based on the following:

quantification. For example, the models. There was never an intent to model the

• the high design temperature limits of the major control room it is not acceptable to not operator actions given that past analyses has shown model a system such as that both rooms can survive a loss of HVAC. It is components, HVAC or CCW on the basis recognized that the analyses requires updating and

• the general conservative modeling assumptions employed throughout that there are procedures for that the documentation requires updating.

the EA-APR-95-023,R1 analysis, dealing with losses of these Palisades should either provide additional justification

• the philosophy of the operators with respect to remaining in the control systems.

for not modeling the HV AC systems for the cable room during such an event, spreading room and control room, or model the

• the TMM's are not credited in the EOPs, operator actions to implement altemate cooling strategies or model HVAC for these two rooms.

• the relative un-importance of HVAC failure on a variety of plant PRA studies.

Therefore it is considered unnecessary to model either loss of HVAC as an initiator or as a support system for the intemal events model.

An analysis of the cable spreading room heat-up following a loss of ventilation was developed using the GOTHIC software code [12]. The analysis approach was to establish the room's heat load based on Systematic Evaluation Program (SEP) Topic IX-5 (Phase II cable spreading room loss of HVAC testing) data by modeling the test boundary conditions in detail and iterating on room heat generation until the test results were mimicked by the model. With the room heat load established, the model boundary conditions were changed to establish a conservative scenario with no room ventilation. The room temperature profile demonstrated that at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> the peak temperature would reach 122°F (50°C).

CALC-455-001-DC2 [13] was then performed to evaluate all cable spreading room equipment modeled in the PRA under these conditions.

The analysiS conservatively assumed that the room was at the peak calculated temperature of 122DF (50DC) for the entire 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> duration of the transient. An evaluation of eqUipment qualification reports, and vendor data, was then performed which concluded that reasonable assurance of operability is assured for all equipment at an elevated ambient of 122DF for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Based on the conclusions of these analyses, ventilation to the cable spreading area is not explicitly modeled as failure to re-establish ventilation does not result in equipment failure prior to the PRA 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time. of NB-PSA-ETSC 111] has been ~dated to reflect these Page 50 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions DO NOT USE proceduralized recovery actions as the sole basis for eliminating a support system from the model; however, INCLUDE these recovery actions in the model quantification. For example, it is not acceptable to not model a system such as HVAC or CCW on the basis that there are procedures for dealing with losses of these systems.

Detailed analysis of systems/component Dependency on HVAC/ventilation should be provided in the individual systems and/or Dependency Tables.

Palisades needs to provide better documentation of the basis for not modeling the HVAC within the system notebooks for the control room and cable spreading room.

SWS pump failures to start are valid contributors to EDG failure. A review of the affected cutsets still has a cutset with a diesel generator run failure in the same cutset as the SWS pump failure to start. Given failure of the SWS pump to start, the diesel generator fail to run should be 1.0. The model needs to account for these and similar dependencies.

These specific failures should be incorporated into the fault tree model. And, given the similarity of this finding with Finding SY-B5-02, it is recommended that a systematic review of other potentially risk important dependencies be performed.

The basis for excluding control room HVAC from the full power internal events model was strengthened to include other aspects in addition to operator actions. This evaluation was fully documented in NB-PSA-ETSC

[11]. The conclusion summary states:

Control room cooling in the Palisades internal events PRA is not considered an issue based on the following:

• the high design temperature limits of the major control room components

• the general conservative modeling assumptions employed throughout the EA-APR-95-023 Rev. 1 analysis

• the philosophy of the operators with respect to remaining in the control room during such an event,

• the TMMs are not credited in the EOPs

• the relative un-importance of HVAC failure on a variety of plant PRA studies Therefore it is considered unnecessary to model either loss of HVAC as an initiator or as a support system for the internal events model.

With respect to cable spreading room cooling. An analysis of the cable spreading room heat-up following a loss of ventilation was developed using the GOTHIC software code [12]. The analysis approach was to establish the rooms heat load based on Systematic Evaluation Program (SEP) Topic lX-5 (Phase II cable spreading room loss of HVAC testing) data by modeling the test boundary conditions in detail and iterating on room heat generation until the test results were mimicked by the model.

With the room heat load established, the model boundary conditions were changed to establish a conservative scenario with no room ventilation.

The room temperature profile demonstrated that at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> the peak temperature would reach 122°F (50°C).

CALC-455-001 -DC2 [13] was then performed to evaluate all cable spreading room equipment modeled in the PRA under these conditions.

The analysis conservatively assumed that the room was at the peak calculated temperature of 122°F (50°C) for the entire 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> duration of the transient. An evaluation of equipment qualification reports, and vendor data, was then performed which concluded that reasonable assurance of SY-B1 2-02 Finding F&O #

(Supporting Finding or ASME Reg. Guide 1.200 Requirement)

Suggestion Category II Text Finding Description (summary discussion)

Disposition conclusions.

Finding Resolved Page 51 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion SY-B12-02 Finding ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review ~eport Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text conclusions.

Finding Resolved DO NOT USE Detailed analysis of systems/component Dependency The basis for excluding control room HVAC from the full power intemal proceduralized recovery on HV AC/ventilation should be provided in the events model was strengthened to include other aspects in addition to actions as the sole basis for individual systems and/or Dependency Tables.

operator actions. This evaluation was fully documented in NB-PSA-ETSC eliminating a support system Palisades needs to provide better documentation of

[11). The conclusion summary states:

from the model; however, the basis for not modeling the HVAC within the Control room cooling in the Palisades intemal events PRA is not INCLUDE these recovery system notebooks for the control room and cable considered an issue based on the following:

actions in the model spreading room.

quantification. For example, SWS pump failures to start are valid contributors to

• the high design temperature limits of the major control room it is not acceptable to not EDG failure. A review of the affected cutsets still has components model a system such as a cutset with a diesel generator run failure in the

• the general conservative modeling assumptions employed throughout HVAC or CCW on the basis that there are procedures for same cutset as the SWS pump failure to start. Given the EA-APR-95-023 Rev. 1 analysis dealing with losses of these failure of the SWS pump to start, the diesel generator

• the philosophy of the operators with respect to remaining in the control fail to run should be 1.0. The model needs to account systems.

for these and similar dependencies.

room during such an event, These specific failures should be incorporated into the

• the TMM's are not credited in the EOPs fault tree model. And, given the similarity of this

• the relative un-importance of HVAC failure on a variety of plant PRA finding with Finding SY 02, it is recommended that studies a systematic review of other potentially risk important Therefore it is considered unnecessary to model either loss of HVAC as dependencies be performed.

an initiator or as a support system for the intemal events model.

With respect to cable spreading room cooling. An analysis of the cable spreading room heat-up following a loss of ventilation was developed using the GOTHIC software code [12). The analysis approach was to establish the room's heat load based on Systematic Evaluation Program (SEP) Topic IX-5 (Phase II cable spreading room loss of HVAC testing) data by modeling the test boundary conditions in detail and iterating on room heat generation until the test results were mimicked by the model.

With the room heat load established, the model boundary conditions were changed to establish a conservative scenario with no room ventilation.

The room temperature profile demonstrated that at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> the peak temperature would reach 122°F (50°C).

CALC-455-001-DC2 [13) was then performed to evaluate all cable spreading room equipment modeled in the PRA under these conditions.

The analysis conservatively assumed that the room was at the peak calculated temperature of 122°F (50°C) for the entire 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> duration of the transient. An evaluation of equipment qualification reports, and vendor data, was then performed which concluded that reasonable assurance of Page 51 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)Q)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

Finding or ASME Reg. Guide 1.200 (Supporting Suggestion Category II Text Finding Description (summary discussion)

Disposition Requirement) operability is assured for all equipment at an elevated ambient of 122°F for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Based on the conclusions of these analyses, ventilation to the cable spreading area is not explicitly modeled as failure to re-establish ventilation does not result in equipment failure prior to the PRA 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time. of NB-PSA-ETSC [11] has been updated to reflect these conclusions.

With respect to SWS pump failure to start, this is not considered a finding.

SWS start failures are captured under the diesel failure to start gates.

Start, load/run and run failures are all captured under OR gates so the logic is equivalent.

The PRA model cutsets properly account for diesel run and service water pump start failures based on evaluation documented in EA-PSA RG1.200F&O-1 0-01 [9].

Finding Resolved SY-B14-01 Suggestion IDENTIFY SSCs that may One potential weakness identified is the Common cause sump blockage events were added to the model (gates be required to operate in documentation and handling of the Containment CCF-316 and CCF-317) and documentation updated (Attachment 13 of conditions beyond their Sump Blockage potential. A discussion of the sump NB-PSA-DA [5]). Description of sump strainer and discussion of sump environmental qualifications, blockage potential was not found in the SSS blockage was added to the SIRWT tank and containment sump suction INCLUDE dependent notebook, and a common cause sump blockage event system notebook (NB-PSA-SY-SSS [21], Sections 1.0, 1.1 and 2.12).

failures of multiple SSCs was not found in the associated fault tree model.

that result from operation in Note: independent sump blockage events are Suggestion Resolved these adverse conditions.

included in the model.

Examples of degraded environments include (a)

LOCA inside containment Because of the significance and impact of the sump with failure of containment blockage potential, the impact of this issue should be heat removal (b) safety discussed in the system notebook and included in the relief valve operability (small model as appropriate. (Note: Palisades did identify LOCA, drywell spray, severe this issue in their self-assessment but it remained accident) (for BWRs) (c) unresolved at the time of the peer review.).

steam line breaks outside containment (d) debris that could plug screens/filters (both internal and external to Page 52 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion SY-B14-01 Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text operability is assured for all equipment at an elevated ambient of 122°F for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Based on the conclusions of these analyses, ventilation to the cable spreading area is not explicitly modeled as failure to re-establish ventilation does not result in equipment failure prior to the PRA 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time. of NB-PSA-ETSC [11] has been updated to reflect these conclusions.

With respect to SWS pump failure to start, this is not considered a finding.

SWS start failures are captured under the diesel failure to start gates.

Start, load/run and run failures are all captured under 'OR' gates so the logic is equivalent.

The PRA model cutsets properly account for diesel run and service water pump start failures based on evaluation documented in EA-PSA-RG1.200F&O-10-01 [9].

Finding Resolved IDENTIFY SSCs that may One potential weakness identified is the Common cause sump blockage events were added to the model (gates be required to operate in documentation and handling of the Containment CCF-316 and CCF-317) and documentation updated (Attachment 13 of conditions beyond their Sump Blockage potential. A discussion of the sump NB-PSA-DA [5]). Description of sump strainer and discussion of sump environmental qualifications.

blockage potential was not found in the SSS blockage was added to the SIRWT tank and containment sump suction INCLUDE dependent notebook, and a common cause sump blockage event system notebook (NB-PSA-SY -SSS [21], Sections 1.0, 1.1 and 2.12).

failures of multiple SSCs was not found in the associated fault tree model.

Suggestion Resolved that result from operation in Note: independent sump blockage events are these adverse conditions.

included in the model.

Examples of degraded environments include (a)

LOCA inside containment Because of the significance and impact of the sump with failure of containment blockage potential, the impact of this issue should be heat removal (b) safety discussed in the system notebook and included in the relief valve operability (small model as appropriate. (Note: Palisades did identify LOCA, drywell spray, severe this issue in their self-assessment but it remained accident) (for BWRs) (c) unresolved at the time of the peer reView.).

steam line breaks outside containment (d) debris that could plug screens/filters (both internal and external to Page 52 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

F A

ME G

• d 1 200 (Supporting inding or S

Reg.

Ui e Finding Description (summary discussion)

Disposition Requirement)

Suggestion Category II Text the plant) (e) heating of the water supply (e.g., BWR suppression pool, PWR containment sump) that could affect pump operability (f) loss of NPSH for pumps (g) steam binding of pumps.

INCLUDE operator interface dependencies across systems or trains, where applicable.

SY-B15-01 Suggestion INCLUDE operator interface In NB-PSA-CSS, On p24, there is a statement that The door events are not modeled as a probability per year that the specific dependencies across two human actions are modeled, CSS-Door-167 and door is in the open state, and are not considered human failure events systems or trains, where CCS-Door-1 67B and pointed to Attachment B.

(EA-PSA-CCW-HELB-02-1 7 [22]). Updated Sections 2.6 and 2.7 of the applicable.

Attachment B in turn pointed to a file Entitled CCC component cooling system notebook, NB-PSA-SY-CCS [28], to point to System Human Failure Event Table. This table the correct reference.

contained only one event, CCS-PCMT-POC-0909.

Discussions with Palisades PRA personnel indicate Suggestion Resolved.

that the references on page 24 were old references pertaining to a sensitivity cases on the impact of leaving the CSS doors open during a steam line break.

Typo only.

Palisades needs to clean up these references.

Page 53 of 55 F&O#

Finding or (Supporting Requirement)

Suggestion SY-B1S-01 Suggestion ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions ASME Reg. Guide 1.200 Finding Description (summary discussion)

Disposition Category II Text the plant) (e) heating of the water supply (e.g., BWR suppression pool, PWR containment sump) that could affect pump operability (f) loss of NPSH for pumps (g) steam binding of pumps.

INCLUDE operator interface dependencies across systems or trains, where applicable.

INCLUDE operator interface In NB-PSA-CSS, On p24, there is a statement that The door events are not modeled as a probability per year that the specific dependencies across two human actions are modeled, CSS-Door-167 and door is in the open state, and are not considered human failure events systems or trains, where CCS-Door-167B and pointed to Attachment B.

(EA-PSA-CCW-HELB-02-17 [22]). Updated Sections 2.6 and 2.7 of the applicable.

Attachment B in tum pOinted to a file Entitled CCC component COOling system notebook, NB-PSA-SY -CCS [28], to point to System Human Failure Event Table. This table the correct reference.

contained only one event, CCS-PCMT-POC-0909.

Suggestion Resolved.

Discussions with Palisades PRA personnel indicate that the references on page 24 were old references pertaining to a sensitivity cases on the impact of leaving the CSS doors open during a steam line break.

Typo only. Palisades needs to clean up these references.

Page 53 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

A2.6 REFRENCES

[1]

Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities, Revision 2, 2009.

[2]

Combustion Engineering Owners Group (CEOG)

, Industry Peer Review the Probabilistic Safety Analysis (PSA) against the Combustion Engineering Owners Group (CEOG) PSA checklists, RIE 2000-02, CE-N PSD-1 194-P Task 1037.

[3]

ERIN Engineering and Research Inc., PALISADES GAP ANALYSIS REVIEW AND UPDATE, P0495060007-2711-061215, October 2004.

[4]

From David Finnicum to Bradford Grimmel, RG 1.200 PRA Peer Review Against the ASME PRA Standard Requirements For The Palisades Nuclear Power Plant Probabilistic Risk Assessment, LTR-RAM-II-10-015, March 12, 2010.

[5]

Palisades PSA Notebook NB-PSA-DA Rev. 6, Palisades PSA Data Notebook.

[6]

Palisades PSA Notebook NB-PSA-QU Rev. 3, Quantification Guideline.

[7]

EA-PSA-1 999-010 Rev 0, Palisades PSA Bayesian Update.

[8]

Palisades PSA Notebook NB-PSA-HR Rev. 3, Palisades Human Reliability Analysis Notebook Volume 2 (Pre-Initiator Operator Actions).

[9]

EA-PSA-RG1.200F&O-1 0-01 Rev. 0, Resolution of Reg. Guide 1.200 October 2009 Full Power Internal Events Peer Review Findings and Observations.

[10]

Palisades PSA Notebook NB-PSA-IE Rev. 4, Initiating Event Notebook.

[11]

Palisades PSA Notebook NB-PSA-ETSC Rev. 3, Event Trees and Success Criteria.

[12]

EA-PSA-GOTHIC-CSRHEATUP-09-09 Rev. 0, GOTHIC Cable Spreading Room Heat-Up.

[13]

CALC-455-001-DC2 Rev. 0, Evaluation of Equipment in the CSR when Exposed to Elevated Temperatures for 48 Hours.

[14]

PLP0247-07-0004.01 Rev. 2, Palisades Nuclear Plant Thermal Hydraulic MAAP Calculations.

[15]

CALC-455-001 -DC1 Rev. 0, Survivability of Equipment inside Containment Following a PRA LOCA/MSLB.

[16]

EA-PSA-FPIE-FIRE-12-04 Rev. 0, Palisades Internal Events and Fire PRA Model.

[17]

Palisades PSA Notebook NB-PSA-CC Rev. 1, PSA Model Configuration Control.

[18]

Palisades PSA Notebook NB-PSA-SA Rev. 1, RG 1.200 PRA Self-Assessment Against the ASME PRA Standard Requirements.

[19]

EA-PSA-IE-00-001 0 Rev. 0, Calculation of Initiating Event Frequencies in Accordance with CEOG Standards.

[20]

Not Used

[21]

Palisades PSA System Notebook NB-PSA-SY-SSS Rev. 1, SIRW Tank and Containment Sump Suction System.

[22]

EA-PSA-CCW-HELB-02-1 7 Rev. 0, Evaluation of the Impact of a High Energy Line Break in CCW Room with either Door 167 to 590 Corridor Auxiliary Building or 1 67B to Page 54 of 55 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR SO.SSa RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

A2.6 REFRENCES

[1]

Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities, Revision 2, 2009.

[2]

Combustion Engineering Owners Group (CEOG), "Industry Peer Review the Probabilistic Safety Analysis (PSA) against the Combustion Engineering Owners Group (CEOG) PSA checklists', RIE 2000-02, CE-NPSD-1194-P Task 1037.

[3]

ERIN Engineering and Research Inc., "PALISADES GAP ANALYSIS REVIEW AND UPDATE", P0495060007-2711-061215, October 2004.

[4]

From David Finnicum to Bradford Grimmel, "RG 1.200 PRA Peer Review Against the ASME PRA Standard Requirements For The Palisades Nuclear Power Plant Probabilistic Risk Assessment", LTR-RAM-II-10-015, March 12, 2010.

[5]

Palisades PSA Notebook NB-PSA-DA Rev. 6, "Palisades PSA Data Notebook".

[6]

Palisades PSA Notebook NB-PSA-QU Rev. 3, "Quantification Guideline".

[7]

EA-PSA-1999-010 Rev 0, "Palisades PSA Bayesian Update".

[8]

Palisades PSA Notebook NB-PSA-HR Rev. 3, Palisades Human Reliability Analysis Notebook Volume 2 (Pre-Initiator Operator Actions).

[9]

EA-PSA-RG1.200F&O-10-01 Rev. 0, "Resolution of Reg. Guide 1.200 October 2009 Full Power Internal Events Peer Review Findings and Observations".

[10]

Palisades PSA Notebook NB-PSA-IE Rev. 4, "Initiating Event Notebook".

[11]

Palisades PSA Notebook NB-PSA-ETSC Rev. 3, "Event Trees and Success Criteria".

[12]

EA-PSA-GOTHIC-CSRHEATUP-09-09 Rev. 0, "GOTHIC Cable Spreading Room Heat-Up".

[13]

CALC-455-001-DC2 Rev. 0, "Evaluation of Equipment in the CSR when Exposed to Elevated Temperatures for 48 Hours".

[14]

PLP0247-07-0004.01 Rev. 2, "Palisades Nuclear Plant Thermal Hydraulic MAAP Calculations".

[15]

CALC-455-001-DC1 Rev. 0, "Survivability of Equipment inside Containment Following a PRA LOCAlMSLB".

[16]

EA-PSA-FPIE-FIRE-12-04 Rev. 0, "Palisades Internal Events and Fire PRA Model".

[17]

Palisades PSA Notebook NB-PSA-CC Rev. 1, "PSA Model Configuration Control".

[18]

Palisades PSA Notebook NB-PSA-SA Rev. 1, "RG 1.200 PRA Self-Assessment Against the ASME PRA Standard Requirements".

[19]

EA-PSA-IE-00-0010 Rev. 0, "Calculation of Initiating Event Frequencies in Accordance with CEOG Standards".

[20]

Not Used

[21]

Palisades PSA System Notebook NB-PSA-SY-SSS Rev. 1, "SIRW Tank and Containment Sump Suction System".

[22]

EA-PSA-CCW-HELB-02-17 Rev. 0, "Evaluation of the Impact of a High Energy Line Break in CCW Room with either Door 167 to 590 Corridor Auxiliary Building or 167B to Page 54 of 55

ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) the West Engineered Safeguards Room Open.

[23]

Palisades PSA Notebook NB-PSA-SS Rev. 2, Palisades Safe and Stable States.

[24]

Nuclear Management Company, Update of Palisades CDF Model PSAR2b to PSAR2c, Calculation No. EA-PSA-PSAR2c-06-1 0, Rev. 0, June 2006.

[25]

Palisades PSA Notebook NB-PSA-HR Rev. 4, Palisades Human Reliability Analysis Notebook Volume 1 (Post Initiator Operator Actions).

[26]

Palisades PSA Notebook, Level 2 Notebook.

[27]

Palisades Pressurized Thermal Shock (PTS) Probabilistic Risk Assessment (PRA),

ADAMS Accession number ML042880473, October 6, 2004.

[28]

Palisades PSA Notebook NB-PSA-SY-CCS Rev. 2, Component Cooling System.

[29]

EA-PSA-PZR-SRV-1 2-03 Rev. 0, Update of Palisades PRA Consequential Pressurizer Safety Valve Operation Data and Methodology.

[30]

EA-PSA-LOCA-IE-12-02 Rev. 0, Initiating Event Frequencies for Loss of Coolant Accidents for the Palisades Nuclear Plant Probabilistic Risk Assessment.

[31]

ASME/ANS RA-S-2008a, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME and the American Nuclear Society, December 2008.

[32]

NEI 05-04, Revision 1 (Draft), Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard (Internal Events), Nuclear Energy Institute, November 2007.

[33]

SCIENTECH document 17825-1, Palisades Fire PRA Peer Review to Requirements in Part 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications.

[34]

EA-PSA-INTFLOOD-1 3-06 Vol. 1, Palisades Internal Flooding Analysis for Internal Events PRA Identification of Flood Areas, Flood Sources, and Impacted Components, Rev. 0, December, 2013.

[35]

EA-PSA-INTFLOOD-13-06 Vol. 2, Palisades Internal Flooding Analysis for Internal Events PSA Initiating Event Frequencies for Flooding Events Rev. 0, December2013.

[36]

EA-PSA-INTFLOOD-1 3-06 Vol. 3, Palisades Internal Flooding Analysis for Internal Events PRA Calculation of Core Damage Frequency, Rev. 0, December2013.

[37]

Electric Power Research Institute, Pipe Rupture Frequencies for Internal Flooding PRAs, Revision 3, Technical Report 3002000079, Final Report, April 2013.

[38]

EA-PSA-FLOOD-IE-1 3-02 Rev. 0, Internal Flood Initiating Event Frequencies for the Palisades PRA.

[39]

EA-PSA-RI-ISI-00-INDIRECT ANALYSIS, RI-ISI Indirect Effects Evaluation, June 2000.

[40]

AOP-39 Rev. 0, Palisades Abnormal Operating Procedure, Internal Plant Flooding.

[41]

EN-NE-G-012 Rev. 0, Entergy Engineering Guide, Internal Flooding Analysis Guidelines.

[42]

EN-NE-G-012 Rev. 0, Entergy Engineering Guide, Internal Flooding Analysis Guidelines.

[43]

EA-PSA-INTFLOOD-1 3-06 Rev. 0, Palisades Flooding Analysis for Internal Events PSA, December2013.

Page 55 of 55 ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) the West Engineered Safeguards Room Open".

[23]

Palisades PSA Notebook NB-PSA-SS Rev. 2, "Palisades Safe and Stable States".

[24]

Nuclear Management Company, "Update of Palisades CDF Model - PSAR2b to PSAR2c", Calculation No. EA-PSA-PSAR2c-06-10, Rev. 0, June 2006.

[25]

Palisades PSA Notebook NB-PSA-HR Rev. 4, Palisades Human Reliability Analysis Notebook Volume 1 (Post Initiator Operator Actions).

[26]

Palisades PSA Notebook, "Level 2 Notebook".

[27]

"Palisades Pressurized Thermal Shock (PTS) Probabilistic Risk Assessment (PRA),',

ADAMS Accession number ML042880473, October 6,2004.

[28]

Palisades PSA Notebook NB-PSA-SY-CCS Rev. 2, "Component Cooling System".

[29]

EA-PSA-PZR-SRV-12-03 Rev. 0, "Update of Palisades PRA Consequential Pressurizer Safety Valve Operation Data and Methodology".

[30]

EA-PSA-LOCA-IE-12-02 Rev. 0, "Initiating Event Frequencies for Loss of Coolant Accidents for the Palisades Nuclear Plant Probabilistic Risk Assessment".

[31]

ASME/ANS RA-S-2008a, Standard for Level1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME and the American Nuclear Society, December 2008.

[32]

NEI 05-04, Revision 1 (Draft), "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard (Internal Events)," Nuclear Energy Institute, November 2007.

[33]

SCI ENTECH document 17825-1, "Palisades Fire PRA Peer Review to Requirements in Part 4 of the ASME/ANS Standard for Level1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications".

[34]

EA-PSA-INTFLOOD-13-06 Vol. 1, "Palisades Internal Flooding Analysis for Internal Events PRA - Identification of Flood Areas, Flood Sources, and Impacted Components,"

Rev. 0, December, 2013.

[35]

EA-PSA-INTFLOOD-13-06 Vol. 2, "Palisades Internal Flooding Analysis for Internal Events PSA - Initiating Event Frequencies for Flooding Events" Rev. 0, December 2013.

[36]

EA-PSA-INTFLOOD-13-06 Vol. 3, "Palisades Internal Flooding Analysis for Internal Events PRA - Calculation of Core Damage Frequency," Rev. 0, December 2013.

[37]

Electric Power Research Institute, "Pipe Rupture Frequencies for Internal Flooding PRAs, Revision 3," Technical Report 3002000079, Final Report, April 2013.

[38]

EA-PSA-FLOOD-IE-13-02 Rev. 0, "Internal Flood Initiating Event Frequencies for the Palisades PRA".

[39]

EA-PSA-RI-ISI-OO-INDIRECT ANALYSIS, RI-ISI Indirect Effects Evaluation, June 2000.

[40]

AOP-39 Rev. 0, Palisades Abnormal Operating Procedure, "Internal Plant Flooding".

[41]

EN-NE-G-012 Rev. 0, Entergy Engineering Guide, "Internal Flooding Analysis Guidelines".

[42]

EN-NE-G-012 Rev. 0, Entergy Engineering Guide, "Internal Flooding Analysis Guidelines".

[43]

EA-PSA-INTFLOOD-13-06 Rev. 0, "Palisades Flooding Analysis for Internal Events PSA," December 2013.

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