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
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 t Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Ml 49043-9530 Covert, MI 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 Proposed Alternative, Use of Alternate ASME Code Case N-716 Palisades Nuclear Plant Docket 50-255 License No. DPR-20

Dear Sir or Madam:

(ENO) hereby requests Pursuant to 10 CFR 50.55a(a)(3)(i), Entergy Nuclear Operations, Inc. (END)

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 (ISI) program. The information provided in the enclosed informed, safety-based inservice (lSI) 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 END 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

201 4-074 PNP 2014-074 Page 2 of 2

Enclosure:

Proposed Alternative cc: Administrator, Region III, Ill, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC

ENCLOSU RE ENCLOSURE ENTERGY NUCLEAR ENTERGY OPERATIONS, INC.

NUCLEAR OPERATIONS, INC. (ENO)

(ENO)

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

Alternative Provides Alternative Provides Acceptable Acceptable Level Level of of Quality Quality and Safety and Safety Risk-Informed/Safety Based Risk-Informed/Safety Based Inservice Inspection Alternative Inservice Inspection Alternative for for Class and 22 Piping Class 11 and Piping 30 pages 30 follow 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-Site - Palisades Nuclear Power Plant (PNP)

Unit:

Interval Interval-- 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 11 and 2 piping welds - Examination Categories B-F, B-J, C-F-1, C-F-i, Components and C-F-2.

Affected:

th 4

Applicable For PNP, the 4th (ISI) Interval began on ten year Inservice Inspection (lSI)

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.

Components,2001 Applicable The requirements from which an alternative is requested are specified in Code the ASME Code,Section XI, Xl, 2001 Edition with the 2003 Addenda, Requirements: IWB-2200, IWB-2420, IWB-2430, IWB-2500, Table IWB-2500-1, 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).

C-F-1 Reason for The objective of this submittal is to request the use of a risk-Request: (RISB) lSI informed/safety based (RIS_B) ISI process for the inservice inspection of Class 1 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 (END),

and Basis for ASME Section XI Xl lSI ISI program for Class 11 and 2 piping. The RIS_B Use: process used in this submittal is based upon ASME Code Case N-716, N-7i6, Alternative Piping Classification and Examination Requirements, Section Xl, Division 1.

XI, 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 1 of 30

ENO, PALISADES NUCLEAR POWER PLANT 50.55a RELIEF REQUEST NUMBER RR 4-20 10 CFR SO.SSa PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i) 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 1.178, An Approach For Plant-Specific Risk-Informed Basis and RG 1.17B, 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 InspectionlNDE 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 1.178.

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 1, 2, 3, or Non-Class). The flooding analysis was performed in (Class 1,2,3, accordance with RG 1.200 and ASMEIASME/ANS ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-200B, ASMEIANS Level 1/Large Early Release RA-S-2008, Standard for Level11Large 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 1.178, this proposed RIS_B program will continue to maintain an RG 1.17B, 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 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-2500- 1 (Examination IWC-2420, IWC-2430, IWC-2500, Table IWC-2500-1 Page 2 of 30

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

C-F-i and C-F-2), is requested in accordance with 10 CFR Categories C-F-1 50.55a(a)(3)(i). A Palisades specific template for the application of SO.SSa(a)(3)(i).

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, 201S).

2015).

Alternative:

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

References:

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

100610470.

D. C. Cook Safety Evaluation - See ADAMS Accession No. ML072620553.

ML072620SS3.

Grand Gulf Nuclear Station Safety Evaluation- See ADAMS Accession ML072430005.

No. ML07243000S.

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

ML080980120.

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ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 1010 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 IE Initiating Events Analysis IF Internal Flooding IFIV Inside First Isolation Valve IGSSC Intergranular Stress Corrosion Cracking IILOCA LOCA Isolable Loss of Coolant Accident IPE Individual Plant Evaluation LE Analysis LERF Anaiysis 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

ENO, PALISADES NUCLEAR POWER PLANT 50.55a RELIEF REQUEST NUMBER RR 4-20 10 CFR SO.SSa 5055a(a)(3)(i)

PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(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 Xl Section XI 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

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 151 Programs

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

3. Risk-Informed/Safety-Based lSIISI 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) lSl Plan Change

4. Proposed lSI

5. References/Documentation Attachment A: Consideration of the Adequacy of Probabilistic Risk Assessment Model for Application of Code Case N716 N71 6 Palisades PRA Response to RG 1.200 Peer Review Findings Page 7 of 30

ENO, PALISADES NUCLEAR POWER PLANT 50.55a RELIEF REQUEST NUMBER RR 4-20 10 CFR SO.SS8 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i) 50.55a(a)(3)(i)

1. INTRODUCTION Palisades Nuclear Power Plant (PNP) is currently in the fourth Inservice (ISI) interval, as defined by the American Society of Mechanical Inspection (lSI)

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 ISI interval began on May 13, 2012.

period of the fourth lSI The ASME Section XI Code of record for the fourth lSI ISI interval is the ASME Xl 2001 Edition with the 2003 Addenda for Examination Category B-F,Section XI C-F-i, and C-F-2 Class 1,2,3, B-J, C-F-1, 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 N-716, Xl Division 1, which is founded in large part on the RI-ISI XI 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-716 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 Probabillstic Risk Assessment Results for Risk-Informed Activities, was of Probabilistic 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 Page 8 of 30

ENO, PALISADES NUCLEAR POWER PLANT 50.55a RELIEF REQUEST NUMBER RR 4-20 10 CFR SO.SSa PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i) 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 Probabillstic Risk EPRI Report 1021467, Nondestructive Evaluation: Probabilistic 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 ISI PROGRAMS 2.1 ASME Section XI ASME Section XI Examination Categories B-F, B-J, C-F-1, C-F-i, and C-F-2 currently contain requirements for the nondestructive examination (NDE) of Class 1 1 and 2 piping components.

The alternative RIS_B Program for piping is described in Code Case N-716. N-7i6.

The RIS_B Program will be substituted for the current program for Class 1 1 and 2 piping (Examination Categories B-F, B-J, C-F-1 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 XI 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 RISB application scope (Le.,

piping with the RIS_B (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 89-i 3, Service Water System Problems Affecting Generic Letter 89-13, 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>i1E-6 based on internal flooding results). While the sampling (CDF>

percentages of Code Case N-716N-71 6 will be applied to this piping, it will be inspected under the existing effective localized corrosion program, TR-i 12657.

per Section 3.6.7 of EPRI TR-112657.

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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) 50.55a(a)(3)(i)

• A plant augmented inspection program has been implemented at PNP in (1),

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 (1)(2)

Case N-770-1 (1)(2).

Notes:

21, 2011, 10 CFR SO.SSa (1) Effective July 21,2011, 50.55a was amended via rulemaking to incorporate by reference Code Case N-770-1,N7701, Alternative Examination Requirements and Acceptance Standards for Class 1 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 W86 182 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 82/1 82 welds per the requirements of Code Case N-770-1.

(2) Alloy 821182 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-146, 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 RISB Program.

incorporated into the RIS_B

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

in Code Case N-716

• 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) 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) 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 ISI Program were used to define the piping system boundaries. Per Code Case plant lSI 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, 2, 3, or Non-Class welds.

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

(2) Applicable portions of the shutdown cooling pressure boundary function. That is, Class 1 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 (i.e., farthest from the RPV) capable of remote closure or to isolation valve (Le.,

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 (i.e., farthest from the RPV) capable of remote closure or to the valve (Le.,

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, N-71 6, this may include Class 3 or Non-Class piping.

(5)

(S) Any piping segment whose contribution to core damage frequency (CDF) is greater than 1 1 E-06 [and per NRC feedback on the Grand Gulf and D. C. Cook RIS_B applications 1 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 TR-1 12657 (Le.,

(i.e., the EPRI RI-ISI methodology), with the exception of the deviation discussed below.

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ENO, PALISADES NUCLEAR POWER PLANT 10 CFR 50.55a RELIEF REQUEST NUMBER RR 4*20 4-20 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a{a){3){i) 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-112657TR-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 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 ~T assumed equal to the greatest potential ~ ATT 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 TASCS, 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 TASCS ASCS susceptibility criteria is presented below.

• Turbulent Penetration TASeS 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 ATs can develop in lines that then turn horizontal, significant top-to-bottom cyclic ~Ts the horizontal sections if the horizontal section is less than about 25 pipe diameters from the reactor coolant piping. Therefore, T TASCS ASCS is considered for this configuration.

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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) 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 ~TsATs 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 ~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 TASCS ASCS will not be significant under these conditions and can be neglected.

• Low flow T TASCS 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) (TT) will govern.

• Valve leakage TASCS Sometimes a very small leakage flow of hot water can occur outward past a significant temperature valve into a line that is relatively colder, creating a Significant steady-state phenomenon with difference. However, since this is generally a "steady-state" no potential for cyclic temperature changes, the effect of T TASCS 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 TTASCS 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 TASCS 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 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) 50.55a(a)(3)(i) program (MRP) MRP-146 guidance on the subject of T TASCS ASCS was also incorporated into the PNP RIS_B application.

3.3 Element and NDE Selection Code Case N-716 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 250/025% 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 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 10% %

of the HSS weld population, additional welds shall be selected so that the total number selected for examination is at least 10 10%.

(2) At least 110%

00/0 of the RCPB welds shall be selected.

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

the RPV.

(4) A minimum of 100/010% 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 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 TR-112657 these locations. Only those RIS_B inspection locations that receive a volumetric examination are included.

I' Class 11 Welds(

)

5 H

1 Welds(1)(5) Class 2 Welds(2) 2 Welds J All Piping Welds(3)(4) 34 Welds Unit Total Selected I'" Total Selected Total Selected 11 626 72 885 19 1523 96 Notes:

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

5.

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

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

1, 2, and 3 ASME Section XI (3) Regardless of safety significance, Class 1,2, 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.

82/182 welds susceptible to no degradation mechanism (5) As described in Section 2.2, Alloy 821182 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 Xl inspection methodology ISI examination of Class 1 for lSI 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 IWB-351 4, Standards for Examination Category 8-F, evaluated per IWB-3514, 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 8-J, 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. 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 50.55a RELIEF REQUEST NUMBER RR 4-20 10 CFR SO.SSa PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.S5a{a){3){i) 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>>(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 (i.e., the practicality of performing dependent on practical considerations (Le.,

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 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 900/090% coverage is not obtained will be submitted per the requirements of 10 CFR 50.55a(g)(5)(iv).

As discussed in EPRI TR-112657, 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 N-716 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%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 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 cooling system (ECCS) injection lines of PWRs, the piping emergency core COOling 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% 90%

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

coverage of a less susceptible location.

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

For locations with no identified degradation mechanisms (Le., (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-716, 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-716, 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 N-71 6 has adopted the NRC approved EPRI TR-112657 Code Case N-716 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 1.17B.1.178. Section 3.7.2 of EPRI TR-112657 requires that the cumulative change in CDF and LERF be less than 1 1 E-07 and 11 E-OB E-08 per year per system, respectively.

For LSS welds, conditional core damage probability (CCDP)/conditional large early release probability (CLERP) values of 1 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 1 E-4/1 E-5 values is that the change-in-risk evaluation process of Code Case N-716 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 11 E-4 (CCDP)/1 E-5 (CLERP) and between medium and low consequence categories are 1 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 1 E-5 to 3E-5 due to an update, it will remain below the 11 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 11 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"

("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 N-71 6 that is consistent with the "simplified Code Case N-716 simplified risk quantification method described in Section 3.7 of EPRI TR-112657. The analysis estimates method" 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 CCDP5).

and small LOCA CCDPs).

Also, as described in Section 2.2, Alloy 821182 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)

Break I Estimated Consequence Upper I Lower Bound Location I CCDP I CLERP Rank CCDP CLERP Description of Affected Piping LOCA 9E-03 I 9E-04 (U) 9E-03 (U) 9E-04 The highest CCDP for Large LOCA HIGH Unisolable RCPB piping of all sizes 1 E-04 (L) 1E-04 1 E-05 (L) 1E-05 (IE_LBLOCA)

(IE LBLOCA) was used with 0.1 LERP IILOCA LOCA 3E-05 I 3E-06 Piping between 11st st and 2nd normally Calculated based on LOCA 1 E-04 (U) 1E-04 1 E-05 (U) 1E-05 MEDIUM open isolation valve inside containment (IE_LBLOCA) CCDP and CLERP times 1 E-06 (L) 1E-06 1 E-07 (L) 1E-07 (letdown and charging) valve fail to close probability -3E-3 3E-3 PLOCA 9E-06 I 9E-07 Piping between 11stst and 2nd normally Calculated based on LOCA 1 E-04 (U) 1E-04 1 E-05 (U) 1E-05 MEDIUM closed isolation valve inside (IE_LBLOCA) CCDP and CLERP times 1 E-06 (L) 1E-06 1 E-07 (L) 1E-07 containment (aux spray, ECCS, drains) valve internal rupture -1 1 E-3 PPLOCA <1E-06 I <1E-07 Calculated based on LOCA 1 E06 (U) 1E-06 1 E07 (U) 1E-07 LOW SDC Class 2 piping (IE_LBLOCA) CCDP and CLERP times 1 E-06 (L) 1E-06 1 E-07 (L) 1E-07 2 valves in series internal rupture <1 E-4 MSLB-l MSLB-I 1 E-05 1E-05 I 1 E-06 1E-06 (V) 1E-04 1 E-04 1 E-05 (V) 1E-05 Main Steam and Feedwater line breaks MEDIUM MS and FW piping inside containment 1 E-06 (M) 1E-06 1 E-07 (M) 1E-07 inside containment CCDP and 0.1 LERP MSLB-O 1 E-02 1E-02 I 1 E-03 1E-03 Main Steam and Feedwater line breaks 1 E-02 (U) 1E-02 1 E-03 (U) 1E-03 HIGH MS and FW piping outside containment outside containment CCDP and 0.1 1 E-04 (L) 1E-04 1 E-05 (L) 1E-05 LERP Class22 I

Class 1 E-04 1E-04 I 1 E-05 1E-05 All Class 2 system piping designated as LSS I 1 E-04 (U) 1E-04 1 E-05 (U) 1E-05 MEDIUM low safety significant except for piping Estimated based on upper bound for 1 E-06 (L) 1E-06 1 E-07 (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 RCPB isolation valve a LOCA in this limited piping downstream of the first RCPS times the probability that the valve fails is aa 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 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 Xo 0 and is expected to have a x

value less than 1 1 E-OB.

E-08. Piping locations identified as medium failure potential have a likelihood of 20Xo. . These PBF likelihoods are consistent with 0

20x 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 ISI Edition/i 978 Addendum of ASME Section XI) interval (1977 Edition/197B Xl) program requirements on a "per per system" 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 after (the implementation of the RIS_B program) and no before and "after" same "before" 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.

With POD Credit Without POD Credit S ys em System Delta CDF Delta LERF Delta CDF Delta LERF CVCs - Chemical Volume & Control System -4.86E-09

-4.B6E-09 -4.86E-10

-4.B6E-10 -2.70E-09 -2.70E-10 ESS - Engineered Safeguards System

- 1 .06E-1 0 1.06E-10 1 .06E-1 1 1.06E-11 1 .06E-1 0 1.06E-10 1 .06E-1 1 1.06E-11 FWS - Feedwater System 1 .97E-09 1.97E-09 1 .97E-1 0 1.97E-10 1 .97E-09 1.97E-09 1 .97E-1 0 1.97E-10 MSS - Main Steam System

- -4.50E-1 0

-4.50E-10 -4.50E-1 1

-4.50E-11 -4.50E-1 0

-4.50E-10 -4.50E-1 1

-4.50E-11 PCS - Primary Coolant System

- -1 .78E-08

-1.7BE-OB -1 .78E-09

-1.7BE-09 -7.70E-09 -7.70E-1 0

-7.70E-10 RWS - Radwaste System

- 0.00E+0O O.OOE+OO 0.00E+0O O.OOE+OO 0.00E+00 O.OOE+OO O.OOE+00 O.OOE+OO SWS - Service Water System

- 3.OOE-1 1 3.00E-11 3.OOE-1 2 3.00E-12 3.OOE-1 1 3.00E-11 3.OOE-1 2 3.00E-12 VAS - Vent & Air Conditioning System

- 0.OOE+00 O.OOE+OO 0.OOE+00 O.OOE+OO 0.OOE+00 O.OOE+OO 0.OOE+00 O.OOE+OO Total -2.1OE-08

-2.10E-OS -2.1OE-09

-2.10E-09 -8.75E-09

-S.75E-09 -8.75E-10

-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 BER piping in FWS (B) (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 XI 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 XI 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. 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 systems pressure boundary. Currently, the process for or ruptures in a system's selecting inspection locations is based upon terminal end locations, structural discontinuities, and stress analysis results. As depicted in ASME White Paper Inservice Inspection Requirements for Class 1, 92-01-01, Rev. 1, Evaluation of InseNice 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)

B-J Pressure Retaining Welds, this methodology has been ineffective Category 8-J 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 locations susceptibility to degradation and secondly, an independent each location's assessment of the consequence of the piping failure. These two ingredients locations 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 N-71 6 requires that any piping on a plant-specific basis that has a contribution to CDF of greater than 1 1 E-06 (or 1 E-07 for LERF) be included in the scope of the application. PNP identified 1

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-716N-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 XI 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) 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 1 of N-716.

Welds classified as LSS do not require preservice inspection.

4. PROPOSED lSI ISI PLAN CHANGE PNP intends to start implementing the RIS_B Program during the plant'splants third period of lSl the current (fourth) inspection interval. By the end of second period of the fourth lSI 56% of the piping weld examinations required by ASME Section XI interval, 560/0 Xl had been completed for Examination Categories B-F, B-J, C F-1 F-i and C-F-2, with the remaining 44% of the examinations to be completed during the two refueling outages in the third 440/0 period (1 R23 and 1 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 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 N-7i 6.

inspected in accordance with N-716.

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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 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 lSI IS! Methodology to Break Exclusion Region Programs.

2. EPRI TR-112657, TR-1 12657, Revised Risk-Informed Inservice Inspection Evaluation Procedure, Rev. B-A.

3. ASME Code Case N-716, N-71 6, Alternative Piping Classification and Examination Requirements, Section XIXl 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 Probabillstic Risk Assessment Results For Risk-Informed Activities.

Adequacy Of Probabilistic

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

N-716,

8. USNRC Safety Evaluation for DC Cook Nuclear Plant, Units 1 1 and 2, Risk-Informed Safety-Based lSI ISI program for Class 11 and 2 Piping Welds, dated 28, 2007. See ADAMS Accession No. ML072620553.

September 28,2007.

Probabillstic Risk Assessment

9. EPRI Report 1021467 Nondestructive Evaluation: Probabilistic 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. ML1100610470.

ML 00610470.

Supporting Onsite Documentation N-716 Evaluation for Palisades",

12. Structural Integrity Calculation 0800075.302, "N-716 Palisades, 0.

Revision O.

Degradation Mechanism Evaluation

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

for Palisades," 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)

Table 3.1 Code o e Case CdC ase N -

N-71 6S 716 Safety Significance af ety S*"f Igm Icance 0Determination etermmatlon Safety Weld N-71 6 Safety Significance Determination N-716 System(1) 1 System Significance Count(2) 2 Count RCPB SDC SOC PWR: FW PWR:FW BER 1 E-6 CDF> 1E-6 High Low cves 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 Pcs 12 V

-/ V

-/ V

-/

RWS 8 V

-/ V

-/

SWS 17 V

-/

VAS 4 V

-/

535 V

-/ V

-/

91 V

-/ V

-/ V

-/

Summary 51 V

-/ V

-/

Results 38 V

-/ V

-/

all 12 V

-/ V

-/

Systems 4 V

-/ V

-/ V

-/

167 V

-/ V

-/

625 V

-/

Totals 1523 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 ENO, PALISADES NUCLEARNUCLEAR POWERPOWER PLANT PLANT 10 CFR 10 CFR 50.55a 50.55a RELIEF RELIEF REQUEST REQUEST NUMBER NUMBER RR 4-20 RR 4-20 PROPOSED ALTERNATIVE PROPOSED ALTERNATIVE IN IN ACCORDANCE ACCORDANCE WITH WITH 1010 CFR CFR 50.55a(a)(3)(i) 50.55a(a)(3)(i)

Table 3.2 Table 3.2 Failure Potential Assessment Summarv 2

Thermal Localized Localized Flow Flow Fati ue Fati~ue Stress_Corrosion_Cracking Stress Corrosion Cracking Corrosion Corrosion Sensitive Sensitive System System(1) TASCS TASCS TT IGSCC TGSCC ECSCC IGSCC ECSCC PWSCC PWSCC MIC PIT MIC PIT CC E-C E-C FAC FAC CVCS - Chemical Volume &

CVCS -

Engineered Safeguards ESS - Engineered

& Control System Safeguards System System I

'I" Feedwater System FWS - Feedwater I

MSS - Main Steam System PCS - Primary Coolant System RWS - Radwaste System F

'I" SWS - Service Water System Conditioning System VAS - Vent & Air Conditicming Notes:

Notes:

1. Systems

1. Systems are are described described in Table 3.1 in Table 3.1 Notes.

Notes.

2. A

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

and MSS systems.

Page Page 25 25 of of 30 30

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

Table 3.3: Code Case N716 Selections Weld N716 Selection Considerations SystemU1 System(1) 22)

Count Counr Selections RCPB HSS LSS DMs OMs RCPB RCPB (IFIV)

(lFlV) BER (OC) eves CVCS 3 TT V

./ ./(3) v(3) 3 eves cvcs 31 TT V

./ 0 eves cvcs 156 None None V

./ 6 ESS 15 IGSee IGSCC V

./ 4 ESS 4 None ./

V V(3)

./(3) 4 ESS 168 None V

./ 7 ESS 51 None None 0 ESS 458 N/A 0 ws FWS 38 None 4 ws FWS 8 None None V

./ 1 1

ws FWS 95 N/A 0 MSS 167 None 11 MSS 8 None V

./ 8 MSS 51 N/A N/A 0 pes cs 1 1 TASeS TASCS V

./ V

./ 11 pes PCS 7 TASeS,TT TASCS,TT V

./ V

./ 3 pes pcs 3 TASeS,TT,PWSee TASCS,TT,PWSCC V

./ V

./ 0 pes PCS 18 TT V

./ V

./ 10 pes PCS 9 TT V

./ 7 pes PCS 7 TT,PWSee TT,PWSCC V

./ V

./ o 0

pes PCS 190 None None V

./ V

./ 26 pes pcs 6 None V

./ 0 RWS 6 None V

./ V

./ 11 RWS 2 None V

./ 0 sws SWS 17 17 N/A N/A 0 VAS 4 N/A N/A 0 11 TASeS TASCS V

./ V

./ 11 7 TASeS,TT TASCSTT V

./ V

./ 3 3 TASeS,TT,PWSee TASCS,TT,PWSCC V

./ V

./ 0 21 TT V

./ V

./ 13 40 TT V

./ 7 7 TT,PWSee TT,PWSCC ./

V ./

V 0 15 IGSee IGSCC V

./ 4 200 None None ./

V V

./ 31 332 None None V

./ 13 16 16 None V

./ 9 9

256 None 4 Totals 898 625 96 Notes:

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

Page 26 of 30

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

Table 3.4: Risk Impact Analysis Results Safety Break Failure Potential Inspections CDF_Impact CDF Impact LERF Impact System Significance Location DMs OMs Rank Rank SXI RIS_B RIS_B Delta wIPOD wlPOD wlo POD w/oPOD wlPOD wIPOD wlo POD wlo CVCS High LOCA TT IT Medium 0 3 3 -4.86E-09

-4.B6E-09 -2.70E-09 -4.86E-10

-4.B6E-10 -2.70E-10 CVCS High PLOCMLOCA PLOCAlILOCA TT IT Medium 0 0 0 0.OOE+00 O.OOE+OO 0.OOE+00 O.OOE+OO 0.OOE÷00 O.OOE+OO 0.OOE÷00 O.OOE+OO CVCS High PLOCA/ILOCA PLOCAlILOCA None Low 0 0 0 0.OOE+00 O.OOE+OO 0.OOE+00 O.OOE+OO O.OOE+OO 0.OOE+00 0.OOE+00 O.OOE+OO CVC Total CVCTotal -4.86E-09

-4.S6E-09 -2.70E-09 -4.S6E-10

-4.86E-1O -2.70E-10

-2.70E-1O ESS High PLOCA/ILOCA PLOCAlILOCA IGSCC Medium 0 4 4 -4.OOE-11

-4.00E-11 -4.OOE-11

-4.00E-11 -4.00E-12

-4.OOE-12 -4.00E-12

-4.OOE-12 ESS High LOCA None Low 0 0 0 0.OOE+00 O.OOE+OO 0.OOE+00 O.OOE+OO OOOE+OO O.OOE+OO O.OOE÷00 O.OOE+OO ESS High PLOCAJILOCA PLOCAlILOCA None Low 0 0 0 0.OOE+00 O.OOE+OO 0.OOE+00 O.OOE+OO 0.OOE+00 O.OOE+OO O.OOE+OO 0.OOE+00 ESS High PPLOCA None Low 0 0 0 0.OOE+00 O.OOE+OO 0.OOE+00 O.OOE+OO O.OOE+OO O.OOE+OO 0.OOE+00 Assume ESS Low Class 2 LSS 11 0 -11 1.1OE-1O 1.10E-10 1.1OE-1O 1.10E-10 1.10E-11 1.1OE-11 1.10E-11 1.1OE-11 Medium ESS Total 7.OOE-11 7.00E-11 7.00E-11 7.OOE-11 7.00E-12 7.OOE-12 7.OOE-12 7.00E-12 FWS High MSLB-l MSLB-I None Low 4 4 0 0.OOE+00 O.OOE+OO O.OOE+OO 0.OOE÷00 O.OOE+OO 0.OOE+00 O.OOE+OO 0.OOE÷00 FWS High MSLB-0 MSLB-O None Low 0 11 1 1 -5.OOE-11

-S.00E-11 -S.00E-11

-5.OOE-11 -S.00E-12

-5.OOE-12 -S.OOE-12

-5.OOE-12 MSLB-l Assume FWS Low MSLB-I 16 16 0 -16 1.60E-1O 1.60E-10 1.60E-10 1.60E-11 1.60E-11 Medium FWS Low MSLB-0 Assume MSLB-O 2 0 -2 2.OOE-09 2.00E-09 2.OOE-09 2.00E-09 2.00E-10 2.OOE-10 2.00E-10 2.OOE-1O Medium FWS Total 1.1OE-1O 1.10E-10 1.1OE-10 1.10E-10 1.1OE-11 1.10E-11 1.10E-11 1.1OE-11 MSS High MSLB-0 MSLB-O None Low 10 10 19 19 9 -4.50E-10

-4.S0E-10 -4.S0E-10

-4.50E-1O -4.S0E-11

-4.50E-11 -4.S0E-11

-4.50E-11 MSLB-l Assume MSS Low MSLB-I 0 00 0 0.OOE+00 O.OOE+OO O.OOE-i-OO O.OOE+OO O.OOE-t-O0 O.OOE+OO 0.OOE+00 O.OOE+OO Medium MSS Total -4.50E-1O

-4.50E-10 -4.50E-10

-4.50E-1O -4.50E-11 -4.50E-11 PCS High High LOCA TASCS TASes Medium 0 1 1 1 1 -1.62E-09

-1.62E-09 -9.00E-10

-9.OOE-1O -1.62E-10 -9.00E-11

-9.OOE-11 PCS High High LOCA TASCSTT TASCS,IT Medium 00 33 33 -4.86E-09

-4.B6E-09 -270E-09

-2.70E-09 -4.B6E-10

-4.86E-1O -2.70E-10 TASCS,TT, TASCS,IT, PCS High LOCA Medium 3 00 -3 2.70E-09 2.70E-09 2.70E-09 2.70E-10 2.70E-10 PWSCC PCS High LOCA TT IT Medium 0 10 10 10 10 -1.62E-08

-1.62E-OB -9.OOE-09

-9.00E-09 -1.62E-09

-1.62E-09 -9.00E-10

-9.OOE-10 PCS High PLOCMLOCA PLOCAlILOCA TT IT Medium 0 7 7 -1.26E-10 -7.OOE-11

-7.00E-11 -1.26E-11 -7.OOE-12

-7.00E-12 PCS High LOCA TT,PWSCC IT,PWSCC Medium 22 00 -2 1.80E-09 1.BOE-09 1.BOE-09 1.80E-09 1.80E-10 1.BOE-10 1.BOE-10 1.80E-10 PCS High High LOCA None None Low 30 21 -9 4.05E-10 4.0SE-10 4.0SE-10 4.05E-1O 4.05E-11 4.0SE-11 4.OSE-11 4.05E-11 PCS High High PLOCA/ILOCA PLOCAIILOCA None Low Low 00 0 00 0.OOE+00 O.OOE+OO 0.OOE+00 O.OOE+OO 0.OOE+00 O.OOE+OO 0.OOE+00 O.OOE+OO PCS Total -1.79E-08

-1.79E-oS -7.77E-Q9

-7.77E-09 -1.79E-Q9

-1.79E-09 -7.77E-10

-7.77E-1O Page 27 of 30

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

Safety Safety Break Failure Potential Failure Potential Inspections Inspections CDF_Impact CDF Impact LERF Impact LERFlmpact System System Significance Significance Location DMs DMs Rank Rank SXI RIS_B RIS_B Delta wIPOD wlPOD wlo POD w/POD wlPOD wlo POD RWS RWS High High LOCA LOCA None Low Low 0 0 0 O.OOE+00 O.OOE+OO 0.OOE+00 O.OOE+OO O.OOE+OO O.OOE+OO RWS RWS High High PLOCMLOCA PLOCAlILOCA None Low Low 0 0 0 O.OOE+O0 O.OOE+OO O.OOE+OO O.OOE+00 O.OOE+OO O.OOE+OO RWS Total O.OOE+OO O.OOE÷OO O.OOE+OO O.OOE+OO O.OOE+OO SWS Total Low Assume SWSTotal Class 2 LSS 3 0 -3 3.OOE-11 3.00E-11 3.OOE-11 3.00E-11 3.OOE-12 3.00E-12 3.OOE-12 3.00E-12 Medium Assume VAS Total Low Class 2 LSS 0 0 0 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE÷OO O.OOE+OO O.OOE÷OO Medium Grand Total 81 73 -8 2.30E08

-2.30E-08 1.07E08

-1.07E-Q8 2.30E09

-2.30E-09 -1.07E-09 1.07E09 Notes

1. Systems are described in Table 3.1 Notes.

2. Only those ASME Section XI Xl Code inspection locations that received a a volumetric examination are included in the count.

count. Inspection locations previously subjected to a 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 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",

High, "Medium",

Medium, or "Low" Low depending upon potential susceptibly to the various types of degradation. [Note: Low Safety Significant (LSS) locations were conservatively assumed to be aa rank of Medium (i.e., "AssumeAssume Medium")

Medium)

5. LSS designation is used to identify those Code Class 2 The "LSS" 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

ENO, PALISADES ENO, PALISADES NUCLEAR NUCLEAR POWERPOWER PLANTPLANT 10 CFR 10 CFR SO.SSa 50.55a RELIEF RELIEF REQUEST REQUEST NUMBER NUMBER RR RR 4-20 4-20 PROPOSED ALTERNATIVE PROPOSED ALTERNATIVE IN IN ACCORDANCE ACCORDANCE WITH WITH 10 10 CFR CFR 50.SSa(a)(3)(i) 50.55a(a)(3)(i)

Table 4: Inspection LLocation Sel Tabl Selections C Comparison Safety Potential Failure Potential Significance Break Break Failure Code Code Weld Weld Section XI Section Xl Code Code Case Case N716 N716 S stem System Significance Location Location Category Category Count Count High Low DMs Rank Rank RIS_B High Low OMs Vol Vol Surface Surface RIS_B Other Other CVCS CVCS V 0/ LOCA lOCA TT Medium Medium B-J B-J 33 00 00 33 00 CVCS CVCS V 0/ PLOCA/ILOCA PlOCAlllOCA TT Medium Medium B-J B-J 31 31 00 33 0 00 CVCS CVCS V 0/ PLOCA/ILOCA PlOCAlllOCA None None Low low B-J B-J 156 156 00 11 11 0 66 ESS ESS V 0/ PLOCA/ILOCA PlOCAlllOCA IGSCC IGSCC Medium Medium B-J B-J 15 15 00 0 44 00 ESS ESS V 0/ LOCA lOCA None Low low B-J 4 0 11 0 4 ESS V 0/ PLOCA/ILOCA PlOCAlllOCA None None Low low B-J B-J 168 168 00 10 10 00 77 ESS ESS V 0/ PPLOCA PPlOCA None Low low C-F-i C-F-1 51 00 0 0 00 ESS V 0/ Class 2 LSS Class2lSS Medium Assume Medium C-F-i C-F-1 458 11 11 ii 0 00 FWS FWS V 0/ MSLB-l MSlB-1 None Low low C-F-2 28 4 0 4 0 FWS FWS V 0/ MSLB-O MSlB-O None Low low C-F-2, Aug 18 0 0 11 0 FWS V 0/ MSLB-l MSlB-1 Assume Medium C-F-2 76 16 0 0 0 FWS FWS V 0/ MSLB-O MSlB-O Assume Medium C-F-2 19 2 0 0 0 MSS MSS V 0/ MSLB-O MSlB-O None Low low C-F-2, Aug 175 10 0 19 0 MSS MSS V 0/ MSLB-I MSlB-1 Assume Medium C-F-2 51 0 11 0 0 PCS V 0/ LOCA lOCA TASCS Medium B-J 11 0 0 11 0 PCS V 0/ LOCA lOCA TASCS,TT Medium B-J 7 0 0 3 0 PCS PCS V 0/ LOCA lOCA TASCS,TT, PWSCC Medium B-J 3 3 0 0 0 PCS V 0/ LOCA lOCA TT Medium B-J 18 18 0 10 10 10 10 0 PCS PCS V 0/ PLOCMLOCA PlOCAlllOCA TT Medium B-J 99 0 0 7 0 PCS PCS V 0/ LOCA lOCA TT,PWSCC Medium B-F, B-J 7 2 1 1 0 0 PCS PCS V 0/ LOCA lOCA None Low low B-J 190 30 10 10 21 55 PCS PCS V 0/ PLOCA/ILOCA PlOCAlllOCA None None Low low B-J 6 0 1 1 0 0 RWS RWS V 0/ LOCA lOCA None None Low low B-J 66 00 00 00 1 1

RWS RWS V 0/ PLOCAIILOCA PlOCAlllOCA None None Low low B-J 2 00 00 00 00 SWS SWS V 0/ Class 22lSS LSS Assume Medium Medium C-F-2 17 17 3 00 0 0 VAS VAS V 0/ Class 2 LSS Class2lSS Assume Medium Assume Medium C-F-2 44 00 00 00 00 Totals Totals 1523 1523 81 81 59 73 23 23 Notes Notes 1.

1. Systems Systems are described in are described Table 3.1 in Table 3.1 Notes.

Notes.

Page 29 29 of of 30 30

ENO, PALISADES ENO, PALISADES NUCLEARNUCLEAR POWER POWER PLANT PLANT 10 CFR 10 50.55a RELIEF CFR 50.55a RELIEF REQUEST REQUEST NUMBERNUMBER RR 4-20 RR 4-20 PROPOSED ALTERNATIVE PROPOSED ALTERNATIVE IN ACCORDANCE WITH IN ACCORDANCE WITH 1010 CFR CFR 50.55a(a)(3)(i) 50.55a(a)(3)(i) 2.

2. The column labeled The column labeled "Other" Other is generally used is generally used to to identify identify plant plant augmented augmented inspection inspection program program locations locations credited credited perper Section Section 44 of of Code Code Case Case N-716.

N-716. Code Code Case N-716 Case N-716 allows the existing plant the existing plant augmented inspection program inspection program for for IGSCC (Categories B IGSCC (Categories B through G) G) inin aa BWR BWR to be be credited credited toward toward the 10%10%

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

in risk impact assessment).

3.

3. The failure potential rank The rank for high safety for high safety significant (HSS)

(HSS) locations locations is assigned assigned asas "High",

High, "Medium",

Medium, or "low" Low depending depending upon potential potential susceptibly susceptibly to the various types of degradation. [Note: [Note: low safety significant Low safety significant (LSS)

(LSS) locations were conservatively conservatively assumed assumed to be be aa rank of of Medium Medium (Le.,(i.e., "Assume Assume Medium").

Medium).

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

not included

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

Page Page 30 30 ofof 30 30

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

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 A2.S Full Power Internal Events Peer Review ............................................................................. 5 A2.6 Refrences .................*.......*..............................................*.................................................54 Page 1 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} 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(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 1 - Initial in-process peer review performed January 18, 2010

• Phase 2 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 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 category II in all Regulatory Guide 1.200 standard supporting requirements with at least capability category" some exceptions.

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

8 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 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 DG-1161 guidance on self-assessments to determine the adequacy of PRAs.

This assessment reviewed the peer review facts against the guidance in DG-1122 DG-i 122 and produced a list of recommended actions to address "gaps" 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 B level facts and observations (F&Os) from the peer review certification report. DG-1122 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-1 A-i of DG-1122).

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 8-4 B-4 of DG-1122.

This was the method used in the Palisades Gap Analysis.

A2.4-i lists the recommended actions identified by this evaluation. In general, the Table A2.4-1 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 A2.4-1 Gap Analysis Recommendations Applicable SR Model Changes tern Item Description of Issues Disposition Numbers Needed?

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

Walkdowns/interviews with operators and engineers Walkdownslinterviews have been conducted in the past, but need to be done SY-A4, SY -A4, IIF-B3, F-B3, B No Addressed again in light of recent PRA updates and staffing IF-C8, IF-E8 changes.

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

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

Concurrent unavailabilities should be included in the D 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 F IF-C7 No Addressed 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 G lF-D5a IF-DSa Yes Addressed 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 II IE-Cl 1 IE-C11 No Addressed high flow/dP to isolate ISLOCA was not credited.

Document the rationale for this exclusion.

The pre-initiators were identified primarily based on test Inspection activities also and maintenance activities. Inspection J HR-Al HR-A1 No Addressed should be addressed explicitly for potential pre-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 DG-l 161. Document how the pre-initiator version of DG-1161.

HEPs account for the quality factors noted in DG-1161.

DG-ll 61.

EDG repair is the only case where repair is credited.

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

Barrier availability was generally not accounted for but N

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) 00 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 P QU-E4 No Addressed sensitivity analyses in EA calculations in the documentation of the current version of the model and subsequent updates.

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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 05-04 [32].

Energy Institute (NEI) OS-04 The report summary states:

The ASME PRA Standard (ASMEIANS liThe (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 Best Practices" Practices (SY-A 13-02 andASA9-01),

and Suggestions and fifty-two IIFindings".

ASA9-01), twenty-six "Suggestions" Findings. All of the new F&Os are 4-12.

presented in Table 4-12."

The report concludes:

Overall, the Palisades PRA was found to substantially meet the ASME PRA Standard at IIOverall, Category!!

Capability Category risk-hformed applications. Dependent

/I and can be used to support risk-informed upon the specifics of the application, additional supporting analyses may be needed, particularly for applications that impact elements with unresolved findings or where an application.

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.

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 QU-Ci -01 and HR-G7-01 related to human error dependency analysis have not been completed. Finding QU-D1-01, 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, ENO, PALISADES PALISADES NUCLEAR NUCLEAR POWER POWER PLANT PLANT 10 10 CFR CFR 50.558 50.55a RELIEF RELIEF REQUESTREQUEST NUMBER NUMBER RR RR 4-20 4-20 PROPOSED PROPOSED ALTERNATIVE ALTERNATIVE IN IN ACCORDANCE ACCORDANCE WITH WITH 10 10 CFR CFR 50.55a(a)(3)(i) 50.55a(a)(3)(i)

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

F&O# Finding or Finding or ASME Reg.

ASME Reg. Guide Guide 1.200 1.200 (Supporting (Supporting Finding Description Finding Description (summary (summary discussion) discussion) Disposition Disposition Suggestion Suggestion Category IIII Text Category Text Requirement)

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

applicable by using the applicable additional found) - key word searches on the LER LER In the current version of the flooding PRA, industry data applied to develop In requirements in in 2-2.1. database are not not comprehensive (this appears appears to bebe utilities did not provide any key maintenance contribution to flood frequency is based the maintenance based on flood events because some utilities key words LERs, or the key words provided are on their LERs, are not documented from 1970 1970 - 2011 in the PIPExp database as described in EA-PSA-FLOOD-lE-13-02 Rev. 0, "Intemal EA-PSA-FLOOD-IE-13-02 9ntemal Flood Initiating Event consistent or comprehensive).

Frequencies for the Palisades PRA*

Frequencies 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 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' Palisades specific LER5 were missed.

applicable LERs data.

Since the population is so small, missing even 11 LER Finding Resolved internal-flood frequency.

has an impact on the intemal-flood 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- Internal-flooding LERs (would be very time-intensive) to ensure completeness.

OR A number number of utilities calculate the intemal-flood internal-flood frequency based on the EPRI TR-1 01341 report (note TR-101341 newer report is to be issued imminently). This a newer approach could be approach be used for for Palisades.

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

factor of assumed. frequency frequency and and its its subsequent subsequent use to determine use to determine the maintenance each each flood scenario scenario group group contribution contribution to to flood flood frequency.

The 75%

The capacity factor was 75% capacity was stated stated to be assumed to be assumed by using the applicable by using applicable based requirements based on industry operating on industry data from 1987-operating data 1987- 1995 1995 (as (as An An assumed assumed plant plant capacity capacity factor factor was was not not applied applied inin the the current current flooding requirements in in 2-2.1.

2-2.1. reported in NUREG-CR15750).

reported in NUREG-CRl5750). Since only covers Since this only update.

update. Plant Plant availability to establish availability to establish the the industry industry prior prior for for maintenance maintenance aa subset the years contained subset of the contained in in the LER LER review, review, induced induced flooding flooding is based on is based 3,554 reactor on 3,554 reactor calendar calendar years of of experience experience and since the later and since later years (where (where capacity factors for for from from all all reactors reactors described described in in the PlPExp PIPExp database.

database. The The database database isis industry were higher) industry were higher) areare not not being being included, included, the the described described in in EA-PSA-FLOOD-IE-13-02 EA-PSA-FLOOD-IE-13-02 Rev. Rev. 00 [38]

[38] and and EPRI EPRI Pipe Pipe capacity capacity factor is not reflective of is not of the the actual actual operating operating Rupture Frequencies Frequencies for for Internal Internal Flooding Flooding Probabilistic Probabilistic Risk Risk Assessments Assessments history, and history, and appears appears to be under-estimated.

to be under-estimated. Note: Note: aa Page 66 of Page of 55 55

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

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

F&O# Finding or Finding ASME Reg. Guide 1.200 (Supporting 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 AAl1.4-4

.4-4 information calculated a capacity factor Finding Resolved NUREG-CR/5750).

>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.

IFEV-A5-03 IFEV-AS-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 life of Palisades. The generic priors is based for the "life" 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 overlap of data is acceptable from a In the current update, Palisades specific flood events were excluded from requirements in 2-2.1. why the "overlap" Bayesian updating perspective.

perspective, the prior data for the Bayesian update as documented in Attachment 2 of EA-PSA-INTFLOOD-l3-06 Volume 2 Rev. 0 [3S].

EA-PSA-INTFLOOD-13-06 [35].

Bayesian updating principles require the priors to be independent of the update data. Remove the Finding Resolved "independent" Palisades data from the generic priors, or only use Palisades data since 2002. I IFEV-A5-04 IFEV-AS-04 Finding DETERMINE the flood- LER Screening criteria A Al1.3.1

.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 cooling systems were not considered since generator COOling 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 these systems at power also require the systems to documented from 1970 - 2011 in the PIPExp database as described in be in operation, events associated with these systems EA-PSA-FLOOD-lE-13-02 Rev. 0, "Internal EA-PSA-FLOOD-IE-13-02 Internal Flood Initiating Event Frequencies for the Palisades PRA"PRA [38]. This database is also used in should not be excluded (maintenance events could -

and probably are - the most likely source of potential the latest EPRI guidance for establishing maintenance contribution to flooding (Pipe Rupture Frequencies for Intemal Internal Flooding Probabilistic Probabilistic Risk flooding events associated with these systems).

Assessments Revision 3, 2013, Report 30020000079 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 -

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

conservative.

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

IFEV-A7-0l IFEV-A7-01 Finding INCLUDE INCLUDE consideration of Palisades calculates the human-induced floods during This finding is related to This to the development of the overall plant flood human-induced floods maintenance by by back-calculating "back-calculating" the maintenance- frequency andand its its subsequent subsequent use toto determine determine the the maintenance during maintenance maintenance through induced floods by by taking the overall internal-flood contribution to to flood frequency.

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

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

F&O#

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

Finding Description (summary Disposition Requirement) 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 PRAA "Intemal PRA and 1990s.

early 1990's. EA-PSA-INTFLOOD-1 3-06 Volume 0 Rev.

documented in Attachment 2 of EA-PSA-INTFLOOD-13-06 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 1990s). During that time during the IPE days (early 1990's).

purposes of determining the total initiating event frequency. No credit is I sites maintenance practices included a frame, most sites' 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-internal-flood frequency, it has a direct impact on the intemal-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 1 E-09/yr, The current analysis evaluates truncation down to 11 E-11 E-1 1 to ensure no if the product of the sum of it is potential that 11 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

< 1 (FOl - E sfgrd,

- documented in the main report of EA-PSA-INTFLOOD-13-06 documented EA-PSA-INTFLOOD-1 3-06 Vol. 3 [36].

scenarios for the area, and . and F06 - Aux Bldg) may be artificially Ascreened" 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, itit could exceed the 1 1 E-9/yr CDF if bounding CCDP is bounding CCDP is the each of the scenarios were were in the 2E-1 0/yr range.

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

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

screened.

Lower the truncation truncation limit used during quantification.

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

7.3.2 of 03(03), All FPIE HEPs that have have applicability to to the the flooding model were were re re-in the internal intemal flood Palisades states Human "Human errors developed as as a part screened screened and appropriate appropriate shaping shaping factors were applied applied specifically specifically for scenarios, INCLUDE the of the the internal intemal events PRA have have been been left left at at their flood initiating initiating events.

events. These events were also screened for were also for operator operator following scenario-specific existing failure probabilities.

probabilities. This This would be reasonable accessibility accessibility (Attachment 11 of [36]) to ensure ensure the action could still be be impacts on PSFs PSFs for control given plant response to given that plant to transient and and loss of of completed whenwhen considering water water propagation propagation and and submergence of of Page 88 of 55

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

Table A2.5-1, March Table A2.S-1, March 20102010 Full Full Power Power Internal Internal Events Events Peer Peer Review Review Report Findings and Report Findings and Resolutions Resolutions F&O #

F&O# Finding or Finding or ASME Reg.Reg. Guide Guide 1.200 1.200 (Supporting (Supporting Finding Description Finding Description (summary (summary discussion) discussion) Disposition Disposition Suggestion Suggestion Category IIII Text Category Text Requirement)

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

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

event.

used: (a) additional (a) additional performance shaping factors (PSFs) performance (PSFs) in in the internal internal workload and and stress (above (above events PRA may not PRA may 1 As part appropriate.

not be as appropriate. part ofof Finding Resolved Finding Resolved workload Q that for that similar sequences for similar Palisades did not change the their quantification, Palisades not caused not caused by internal HEP5 from the HEPs the internal internal events HFEs. Therefore, events HFEs.

(b) cue availability (c) floods) (b) Palisades did Palisades did not not address the the flood specific impacts effect of effect flood on of flood on mitigation, mitigation, on the on PSF5.

the PSFs.

required response, required response, timing, timing, and recovery activities (e.g., impacts are such that the HEPs The flood specific impacts HEPs accessibility restrictions, restrictions, carried over are non-conservatively low.

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

training (e.g.,

and training and procedures, training procedures, training exercises) exercises)

IFQU-A7-01 IFQU-A7-01 Finding Finding PERFORM internal PERFORM internal flood flood A truncation limit of 1 1 E-09/yr was used for the Internal The current analysis evaluates truncation down to 11 E-11 E-1 1 to ensure no sequence quantification 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 accordance provided, limit was not provided. are met. The truncation study is documented in the main report of applicable requirements applicable requirements EA-PSA-INTFLOOD-13-06 Vol. 3 [36].

described in in 2-2.7 2-2.7 sufficiently There is no evidence that this truncation is suffiCiently described low to meet the requirements of QU-S3 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 IFQU-A9-01 Finding Finding INCLUDE, INCLUDE, in in the A specific discussion of jet jet impingement and and pipe Added discussion to SectionSection A3.5 A3.S in Attachment 3 of of EA-PSA-INTFLOOD EA-PSA-INTFLOOD-quantification, quantification, bothboth the whips was notnot identified.

identified. 13-06 13-06 Volume 1 1 Rev. 0 [35] regarding how pipe whip whip and jet jet impingement impingement direct direct effects effects of the flood were evaluated evaluated during the ISI lSI walkdowns; walkdowns; as the RI-ISI RI-ISI report provides provides the Consideration of jet impingement and pipe whips (as (e.g.,

(e.g., loss of cooling from a basis for the indirect basis indirect effects on equipment for each flood or spray initiator.

appropriate) are a requirement of the standard for this service service water water train train due to anan The RI-ISI RI-ISI walkdowns walkdowns were not not limited in in time or scope due to escort escort associated element.

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

indirect indirect effects effects such such as as Provide Provide a discussion discussion of howhow jet jet impingements and and Detailed Detailed walkdowns of all all flood flood areas were documented documented in in the RI-ISI submergence, submergence, jet jet pipe whips were pipe whips were considered and and handled.

handled. The indirect effects effects report EA-PSA-lSl-00-INDIRECT EA-PSA-ISI-OO-INDIRECT Rev. Rev. 00 [39].

[39].

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

applicable. walkdowns walkdowns performed performed for for the IPE.

I PE. The The scope scope of of these these Attachment Attachment 55 of of EA-PSA-INTFLOOD-13-06 EA-PSA-INTFLOOD-13-06 Vol. Vol. 11 Rev.

Rev. 00 [34]

[34] to to walkdown was walkdown was limited as a result limited as result of of time time constraints constraints demonstrate additional walkdowns demonstrate additional walkdowns were were performed performed both both before before and and after after placed placed onon the the walkdown walkdown teamteam by by the the authorized team team Page Page 99 of of 5555

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

Table A2.5-1, March Table A2.S-1, March 20102010 Full Full Power Power Internal Internal Events Events Peer Peer Review Review ReportReport Findings Findings and and Resolutions Resolutions F&O #

F&O# Finding or Finding or ASME Reg.

ASME Reg. Guide Guide 1.200 1.200 (Supporting (Supporting Finding Description Finding Description (summary (summary discussion) discussion) Disposition Disposition Suggestion Suggestion Category IIII Text Category Text Requirement)

Requirement) escort. Palisades escort. Palisades indicated that they indicated that they had performed aa had performed the 2008 the 2008 walkdown walkdown in in which which the the escort escort had limited time.

had limited time.

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

Internal Report.

IFSN-A15-01 IFSN-A 15-01 Finding Finding For each For defined flood each defined flood area The Heater Drain pump suction tank T T-5

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

accumulation. significant height. HBD-13-3 Misc West Drain Tank Significant the walkdown documentation in Attachment 5 of T89A/B From Spool To Condensate Storage Tank T89AIB EA-PSA-INTFLOOD-13-06 Vol. 11 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 justification/basis

- 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.

missed.

Provide basis for determining or assuming insufficient volume.

IIFSN-A17-01 FSN-A 17-01 Finding Finding CONDUCT aa plant CONDUCT 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 identified as a result of time constraints placed indirect effects report EA-PSA-ISI-OQ-INDIRECT EA-PSA-ISI-00-INDIRECT Rev. 0 [39]. Additional accuracy accuracy of of information information on the walkdown team by the authorized authorized team escort. walkdown documentation documentation and and clarification was added to Attachment 5 of obtained fromfrom plant plant His limited limited availability resulted in the walkdown team EA-PSA-INTFLOOD-13-06 EA-PSA-INTFLOOD-13-06 Vol. 11 Rev. 0 [34] [34] to to demonstrate demonstrate additional information information sources sources and and to prioritizing the areas reviewed.

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

Because of the limited limited walkdown time, some rooms located within located within each defined were not flood not walked down, down, therefore the pipe pipe lengths lengths for The The current methodology methodology doesdoes not not rely on the the RI-ISI RI-ISI derived derived pipe pipe failure flood area area (b) (b) these rooms flood/spray/other rooms were not identified. This results not identified. results in in the the frequency data.

data. All pipe failure frequencies were were developed developed inin flood/spray/other applicable nfrequency* for the frequency the rooms relying on on Rl-lSl RI-ISI data data accordance withwith latest latest EPRI EPRI methodology methodology [37].

[37]. All All pipe pipe lengths lengths were were mitigative features mitigative features of the the instead instead of of being being able able to to use use the best best available datadata for for obtained from from plant plant isometric isometric drawings.

drawings.

SSCs located SSCs located within within each pipe/component failurefailure rates.

defined flood defined flood area area (e.g.,

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

(c) The Internal Flooding The Internal Flooding Analysis Analysis Report Report referenced pathways that pathways that could could lead lead to to walkdowns walkdowns performed performed for the IPE.

for the IPE. The The scope scope of of these these Page Page 10 10 of of 55

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Table A2.S-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions F&O# Andingor 1.200 Finding or ASME Reg. Guide 1.200 Finding Description Description (summary (Supporting Finding discussion)

(summary discussion) Disposition Disposition Requement) Suggestion Category II Text Requirement) transport to transport the flood to the flood area area walkdown was limited limited as result of as aa result of time time constraints constraints placed on placed on the the walkdown walkdown team team by the authorized by the authorized team team Palisades indicated escort. Palisades indicated that that they had performed aa had performed more recent complete walkdown, recent complete walkdown, but but that that walkdown was not was not referenced in in the Internal Internal Flooding Flooding Analysis Palisades wants to credit the more Report. If Palisades more recent recent walkdown, they need need toto reference itit in in the Internal Internal Flooding Analysis Report.

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

responses that or operator responses identified.

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

propagation. Required by SA. 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.

IIFSN-A6-01 FSN-A6-01 Finding Finding SSCs identified in For the SSCs Spray effects from chilled water systems pipe failures Further documentation for eliminating the chilled water system as a flood IFSN-A5, IDENTIFY the IFSN-A5, dont seem to be addressed. The basis for elimination don't source is based on plant drawings of the rooms transgressed and design susceptibility susceptibility of each SSC in in as a spray consideration is only documented documented for 1 1 information of the Bus 1 1DD room cooler. This documentation was added to a flood area to flood-induced room - not for all

- all rooms transgressed. Attachment 2 of EA-PSA-INTFLOOD-13-06 EA-PSA-INTFLOOD-13-06 Vol. 2 [35].

failure failure mechanisms. This requirement states that the susceptibility of each Finding Finding Resolved INCLUDE failure by submergence SSC in a flood area to flood-induced failures submergence and spray in mechanisms mechanisms by either submergence or spray are the identification identification process.

process.

included.

EITHER:

EITHER: a) a) ASSESS qualitatively qualitatively the impact impact of of Address Address the potential spray spray effects from chilled water flood-induced mechanisms system pipe failures in all zones transgressed.

that that are not formally are not formally addressed (e.g.,

(e.g., using using the mechanisms mechanisms listedlisted under under Capability Category III III of of this this requirement), by by using using conservative conservative assumptions; assumptions; OR OR b)b) NOTE NOTE thatthat these these mechanisms mechanisms are are not not - - L .. ..

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

F&O# Finding or ASME Reg. Guide 1.200 1.200 (Supporting (Supporting Finding Description (summary discussion) Disposition Requirement) Suggestion Category II Text Requirement) included in the scope of the evaluation.

IIFSO-A4-01 FSO-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 The approach to this is described in reports EA-PSA-FLOOD-IE-13-02 EA-PSA-FLOOD-lE-13-02 failure modes of delineating what the human induced event was.

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) Finding Resolved human-induced understand their full impact on the results or address 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 intemal internal system (c) other events flooding.

resulting in a release into the flood area IFSN-A1 2-01 IFSN-A12-01 Suggestion 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)

(Suggestion) and each flood source, reasonable, it is not clear what criteria from the Entergy Internal Intemal Flooding Guidelines procedure EN-NE-G-012 Section IDENTIFY IDENTIFY the propagation Standard was used used for the various flood areas 5.3.3 [41]. This procedure procedure was added as a reference document to the path from the flood source screened. Because of the multiple screening screening criteria flood area development documentation in Attachment 11 of area area to its its area area of of available, available, specifying the criteria applied applied would be EA-PSA-INTFLOOD-13-06 EA-PSA-INTFLOOD-13-06 Vol. 11[34]. [34].

accumulation.

accumulation. beneficial to ensure ensure no no zones zones were inappropriately screened. Suggestion Suggestion Resolved This may result result in in additional additional zones zones being being able able to be screened, screened, andand would would ensure ensure the zones already already screened were done done in in accordance with the the Standards Standard's requirements.

Specify Specify the the criteria criteria from from the the Standard that that was was applied applied toto screen screen the the various zones from further further consideration.

IFSN-A8-01 IFSN-A8-01 Suggestion Suggestion IDENTIFY IDENTIFY inter-area inter-area Inter-area Inter-area propagation propagation through through the the normal normal flow flow path path Further Further documentation documentation forfor eliminating eliminating the the chilled chilled water water system system as as aa flood flood (Suggestion)

(Suggestion) propagation propagation through through the the from one one area to another area to another viavia drain drain lines lines were were source source isis based based on on plant plant drawings drawings of of the the rooms rooms transgressed transgressed and and design design normal normal flow pathpath from from one one addressed addressed by by the the GOTHIC GOTHIC runs.runs. Areas Areas connected connected via via information information ofof the the Bus Bus 11DD room room cooler.

cooler. This This documentation documentation was was added added toto Page 12 Page 12 ofof 55 55

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

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 Attachment 2 of EA-PSA-INTFLOOD-13-06 Vol. 2 [35] and included I lines; and areas connected valves, pipe and cable penetrations (including cable discussion of the potential flow rate.

I I

via backflow through drain trays), doors, stairwells, hatchways. There didn't didnt I

Suggestion Resolved I

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 IDENTIFY the Bus 1 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 11 D room cooler. This documentation was added to (1)]. INCLUDE: (a)

(1)). consideration as a flood source because it has Attachment 2 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 Final List Of Potential Suggestion Resolved the area that are connected Hazards/Postulated Effects In The Bus 1 1D D Room").

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, verity and document that the elimination IPE, need to verify systems) (b) plant internal criteria used used is still valid, especially since there may sources of of flooding (e.g.,

have been chillers installed in have been in the plant plant that use tanks or pools) located in in Chilled Water since the IPE Chilled Water IPE was performed.

the flood area (c) plant external sources of flooding Provide the criteria/basis criterialbasis for how it was determined 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 is still area through some system valid.

or structure (d) in-leakage in-leakage from other other flood areas (e.g.,

back flow through drains, drains, doorways, doorways, etc.)

AS-A2-01 Finding For each modeled modeled initiating initiating The event trees specify specify the required key safety Notebook Notebook NB-PSA-SS, Palisades "Palisades Safe and Stable States States" [23] was event, IDENTIFY IDENTIFY the key key functions needed to mitigate the initiating initiating event event ofof developed to evaluate and document document the non-core non-core damage end states for safety safety functions that are are interest, but mission time is specified but the mission as 24 specified as 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

hours. all event event trees.

trees. Generalized Generalized flow charts charts were were developed developed to to capture capture all of Page 13 13 of 55 55

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A2.5-1 March 2010 Table A2.S-1, , 201 0 Full Power Internal Events Peer Review Report Findings and Resolutions F&O #

F&O# Finding or ASME Reg.Reg. Guide 1.200 1 .200 (Supporting Finding Description (summary discussion)

Finding Disposition 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"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 [11] [1 1] was revised to ensure event, using the success does not always match the current event tree logic. logic, all event tree headings match the headings described in the notebook criteria defined for each key documentation. Section 4.10 was added to the notebook to describe the For some of the event tree nodes there appears to be safety function (in operation of auxiliary feedwater pump P-8B after battery depletion a documentation mis-match. For example: Section 5.9 accordance with SR SC-A3), heading. Section 5.9 was corrected to agree with the number of required IDENTIFY the systems that of NP-PSA-ETSC, r01 rOl states that 3 of 3 charging IDENTIFY charging pumps as described in the success criteria in Section 5.9.4 for pumps are required for a VSBLOCA, but the success can be used to mitigate the 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 AS-A10-01 Finding In constructing the accident Although the event trees include operator actions A copy of the Human Error Probability (HEP) Post-Initiator Calculation (P- (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" 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, applied. was also provided.

on systems and required operator interactions (e.g., The CAT II requirement to capture and provide HEP5 were assigned to the five Operations Department Operating Crews HEPs sufficient detail detail for significant differences differences in in (10 per crew) for review. Their reviews included

(-10 included ensuring indications, systems initiations or systems or valve requirements associated with systems and/or procedure selection and use, and activity performance performance man-power and alignment) are captured.

Where diverse 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),

event 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 comments were reviewed and proposed resolutions forwarded to Operator action is based on timing for Loss of Main changes the requirements the comment initiator for further comment or acceptance. Initiator Initiator Feedwater (LOMFW). There is no no differentiation for operator operator intervention oror comment acceptance was documented documented by their initialing the HEPHEP between the timing for any of the other initiators, initiators, and the need for other systems, Validation Validation form.

it does does not not appear that the LOFW initiating eventevent isis MODEL MODEL each each separately.

the bounding event event for this this operator operator action.

action. Training Training comments were were similarly dispositioned and and documented.

documented. AllAll The operator validations have been completed and the documents attachedattached to Volume II actions as currently evaluated operator actions evaluated need need toto be of the human reliability analysis analysis notebook (Appendix FF of [8]).

[8]).

be reviewed to ensure they are are bounding "bounding" for all all scenarios scenarios where they are credited. To To meet the CAT Finding Finding Resolved.

IIII requirement,_timing_differences_(and_potentially requirement, timing differences (and potentially Page 14 14 of 55 55

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

F&O#

Finding or Guide 1.200 ASME Reg. Guide 1.200 (Supporting 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 [11] [1 1] to clarify that all sequences and treat Event notebooks. No documentation associated with controlled manual shutdown' transient initiators, including 'controlled 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 Special True in the event tree rules file to logical operators that are set to 'True' events); (d) the operator lnitiatorsv) in the SAPHIRE program that are not Initiators*) 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 HHRA developed into the 'Support Support System to Front-Line System Dependency RA for example FMEA",

FMEA provided in the report is an "example FMEA, these actions; (e) the but the actual FMEA performed is not included or Matrix and 'Support Matrix' Support System to Support System Dependency Matrix'. Matrix. This 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

[10]. The final FMEA results were added as Attachment 12 to the initiating sequence models with plant identifies how the final support system initiators were 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 "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 documentation that the key safety functions or success success criteria criteria defined defined is appropriate appropriate and and adequate adequate for them.

Although the Controlled Manual Shutdown is listed in in the table in Attachment 3, there is no mention of it it in the discussion in in Section 3. A discussion of the event needs to be included in Section 3 similar to how the other transients transients are described. Also need to include include the actual actual FMEA in in the documentation or or provide a valid valid reference for it. it.

AS-C2-02 Suggestion DOCUMENT the processes processes While itit is obvious that the Flag files exist, the A summary summary discussion of the event tree rules (Flag) file was added added to used used to develop accident development development and and review ofof the Flag file used used in Section 3.0 ofof the event event trees and success success criteria criteria notebook notebook NB-PSA NB-PSA-sequences sequences and treat treat SAPH SAPHIRE IRE is is not not included inin the the accident sequence sequence ETSC ETSC [1 1]. In

[11]. In addition, addition, Table Table 3.0-1 was added added which which lists all c>f of the -

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

F&O# Finding or Finding or ASME Reg.

ASME Guide 1.200 Reg. Guide 1.200 (Supporting (Supporting Finding Description Finding Description (summary (summary discussion) discussion) Disposition Disposition Suggestion Suggestion Category IIII Text Category Text Requirement)

Requirement) dependencies in dependencies in accident accnient documentation. The Flag analysis documentation. Flag file does does appear appear to to initiating event initiating event logical logical variables variables set to true true in in the rides file for aa given the rules given including the sequences, including be documented in be documented the Quantification in the Quantification report.

report. However, However, initiating transient event. The initiating The summary references references the the detailed detailed discussion inputs, methods, inputs, methods, and because this file because governs how file govems how the accident sequences sequences developing event for developing event tree rules rules which is is documented in in Section Section 3.0 ofof the the results. For example, this quantified, it is important are quantified, important to ensure ensure the accident quantification notebook, quantification notebook, NB-PSA-QU NB-PSA-QU [6].

documentation typically documentation sequences (especially the support system initiators) includes: (a) the linkage Suggestion Resolved.

includes: SAPH IRE model, that the are handled correctly in the SAPHIRE between the the modeled model is modified correctly for applications, and is is initiating event in initiating event in the maintenance and update of important for long term maintenance Event Analysis Initiating Event the model. To support this, documentation of the Flag section and the section the accident file is an important part of the accident sequence sequence model; (b) the documentation.

success criteria established success established modeled initiating It is recommended that Palisades provide at least a for each modeled event including including the bases for brief discussion of the Flag and provide a link to the event the criteria (Le.,

(i.e., the the system documentation as it exists in the quantification report.

the I capacities required to mitigate the accident and the necessary components the required to achieve required achieve these capacities); (c) a description of the accident progression for each sequence or group sequences (Le.,

of similar sequences (i.e.,

descriptions of the sequence deSCriptions timing, applicable procedural guidance, guidance, expected environmental or environmental phenomenological impacts, phenomenological dependencies between dependencies systems systems andand operator actions, end states, and other pertinent other pertinent information information required to fully establish the sequence of events); (d) (d) the operator operator actions reflected in in the the event event trees, trees, andand the the sequence-specific sequence-specific timing and and dependencies dependencies that that are are traceable to the HRA HRA for these actions; actions; (e)

(e) the interface interface ofof the the accident accident sequence seguence models models withwith plant plant Page 16 Page 16 ofof 55 55

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

F&O#

Finding or Finding or ASME Reg.

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

(summary discussion) Disposition Disposition Suggestion Suggestion Category IIII Text Category Text Requirement)

Requirement) damage states; damage states; (f)

(f) [when

[when sequences are sequences are modeled modeled using single top using aa single top event event fault tree] the tree] manner in the manner in which which the requirements for the requirements for accident sequence accident sequence analysis analysis have been have been satisfied.

satisfied.

DA-A2-01 DA-A2-01 Finding Finding ESTABLISH definitions ESTABLISH definitions of Component boundaries defined Component for some defined for some Palisades Palisades modeling intentionally separates contact pairs, The Palisades PRA modeling SSC boundaries, failure SSC components are are not not consistent with the generic data motors. This breakers etc. from pump motors. This isis the correct method method of modeling modes, and modes, and success success criteria criteria component boundaries for the same component. For equipment to ensure that appropriate qualitative inSights plant eqUipment insights are in manner consistent in aa manner consistent with with example, the Palisades data report states that the realized. This practice was demonstrated during conduct of the Industry corresponding basic 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 definitions in Systems Systems pump breaker, while the corresponding Palisades the development of the Palisades logic modeling that commenced in 1982 Analysis (SY-A5, Analysis (SY-A5, SY-A7, component boundary separates the pump breaker in support of the MSIV SEP issue resolution.

SY-A8, SY SY-A9

-A8, SY through SY-

-A9 through and pump into two separate events, with separate A14 and SY SY-B4) for failure failure In addition, an evaluation was performed using an interim model (PSAR3 A 14 and -84) for failure rates for each. This separation also appears to rates and common cause and common cause Release 2b) to determine the magnitude of the potential conservatism rates propagate to the definition of component boundaries failure parameters, parameters, and and introduced by having separate data and component boundaries for failure for common cause failures.

ESTABLISH boundaries of breaker-pump combinations as well as other components supplied with ESTABLISH unavailability events events in in a Component boundaries need to be consistent to avoid electrical power via breakers. To bound the problem described in finding unavailability consistent with manner consistent with potentially double counting failures. DA-A2-01, a change set was developed with the failure probability for all manner corresponding definitions in 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 Analysis (SY Systems Analysis (SY-A19). zero. The release 2b base model core damage frequency with the Systems -A 19). assign a failure probability of "0* 0 to the breaker and normally applied breaker failure probability was 2.29 E-05/yr. With the totala failure rate to the pump itself -

assign the "total failure probability of all breakers set to zero, the core damagedamage frequency including updating updating the generic data with the total atotal" plant-specific failures (pump and associated breaker reduces 19%

19% to 1.85 1.85 E-05/yr. This change is less than afactorofa factor of two plant-specific 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, conservatism, it is deemed unnecessary is deemed unnecessary to redefine boundary requirements of the standard. IfIf differences differences component component boundaries in in the PRA.

PRA. In In summary, modeling equipment between between component boundaries boundaries defined defined inin the subcomponents such 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, retained, these differences differences andand their bases should should be structures and component reliability.

reliability.

included included inin the PRA documentation.

Finding Finding Resolved.

DA-C7-01 DA-C7-01 Finding Finding BASE number of BASE number of Palisades Palisades used used actual plant procedures actual plant procedures andand All preventive All preventive maintenance activities activities for PRA PRA identified identified components components werewere surveillance tests on surveillance tests on plant plant experience experience to count surveillance to count surveillance tests.

tests. Planned Planned collected collected from from Palisades Palisades' current current equipment equipment database database andand reviewed.

reviewed.

surveillance surveillance requirements requirement~ _ maintenance activities

,---I!!~ntenance are estimated activities are estimated rather rather than than Referring Referring to to sections 5.4 of sections 5.4 of the the Data Data notebook, notebook, NB-PSA-DA NB-PSA-DA [5], [5], Palisades Palisades Page 17 Page 17 ofof 55 55

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

F&O# Finding or ASME Reg.Reg. Guide 1.200 (Supporting Finding Description (summary discussion) Disposition 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 an and actual practice. BASE previous operating experience and not based on equivalency was defined based on the three year data window".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 experience, previous plant experience. assigned. While this number is estimated, it is based on the of 0.5 was assigned."

plant experience.

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. I I

I 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 I of offsite power, loss of Attachment 9 of NB-PSA-IE NB-PSA-lE (Initiating Event models for the treatment of LOOP events. The modeling aspects include I 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).

(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 interactions, the August 2003 damage. In addition to analyzing these interactions, 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 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 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 distribution),

failure until the system or appropriate, uncertainty in the uncertainty analysis using a two stage Bayes'Bayes appropriate.

function is retumed returned to methodology. The occurrence and non-occurrence methodology. non-occurrence of LOOP during the In addition, given given that other long-term long-term LOOP LOOP events events NE NE Blackout at each each plant in in the industry data was considered based on on service.

were screened as as not applicable to Palisades, it is is whether a LOOP actually occurred or or not at each site in the the first stage of of suggested that a sensitivity analysis address the Bayes' process. In the Bayes In the second stage, two models were effect of the screening process. probabilistically combined each having having a defined probability of being true.

One model assumed the north north east Blackout Blackout as a Palisades LOOP and Loss of off-site off-site power is an important risk contributor power is contributor the other counted it as a non-event. Sensitivity studies were performed to and the effect of the screening of longer LOOP LOOP events show the effect of altering altering the probability probability of LOOP from the base case of in in the LOOP recovery analysis have have a significant 25% up to 50%.

50010. Further details details of this analysis are are described inin Section impact on the risk.

5.7 of NB-PSA-IE [10].

Document Document treatment of August 14, 14, 2003 2003 North North Suggestion Suggestion Resolved America LOOPLOOP event event and perform a sensitivity sensitivity study.

study.

DA-Di-Ol DA-D1-01 Finding Finding CALCULATE realistic CALCULATE Bayesian updates of all all plant specific calculations calculations Palisades parameter estimates are are calculated calculated based on on Bayesian Bayesian analysis parameter estimates estimates for used average distributions.

used the industry average distributions. For Category employed with a combination combination of plant specific and generic industry

~------

significant basic events significant basic events II, II, it is is necessary necessary toto update update significant significant basic events sources. Generic industry industry sources sources include NUREG/CR-6928, -

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

F&O# Finding or Reg. Guide 1.200 ASME Reg.

(Supporting Finding Description (summary discussion) Disposition 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 EPRI TR-016780 TR-01 6780 Rev.

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 Attachment 10 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 average prior may have distribution spreads that can specific data to characterize documented maintenance rule failures as listed in Attachment 3 of [5].

overwhelm plant experience data when doing a the parameter value and its 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 informative prior or a prior reflecting plant to plant necessary to combine performed using the BART code which graphically illustrates the prior and variability would allow plant operating experience to evidence from generic and posterior distributions on the same plot. During this process, there were have a larger impact on the resulting posterior mean.

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 I 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.

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 1 E-06 and runrun 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 negligible (Le.,

(i.e., less than a the posterior distribution is factor of 2). In cases where there were no plant failures and the failure reasonable given the Based updated data should be confirmed appropriate.

appropriate, rates were above 1 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 include a were reviewed to confirm the impact from the Bayesian update was provided by the prior prior and the discussion of how the requirements of this SR DA-D4 discussion minimal (i.e.,

(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.prior.

Examples of tests to ensure that the updating is In addition, a comparison was made between the data used in the previous analysis [7] to that used in this update. Failure codes in which accomplished correctly and that the generic parameter there was a measurable difference difference in the Bayesian updated plant-specific estimates data data (e.g., factor greater than 5) 5) were were reviewed inin detail.

detail. A spreadsheet estimates are are consistent consistent analysis for each each Bayesian Bayesian update was performed using the BART code was performed with the plant-specific with plant-specific application which which provides a visual comparison of of the prior prior and and posterior distributions.

distributions.

application include include the the following: (a) The Bayesian Bayesian update process process and and reviews performed are are described described in in (a) confirmation confirmation that_the_Bayesian_updating Section 8.1 ofof NB-PSA-DA NB-PSA-DA [5].

that the Bayesian updating Page 19 19 of 55 of

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

F&O#

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

Finding Disposition 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 DA-DS-01 Finding If modifications to plant Plant-specific failure data collected in the collection Section 4.3, "Plant Plant Modifications",

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 Only applicable plant-specific data can be applied to representative of current Attachment 3 of the document.

document.

the failure events.

performance, LIMIT the use of old data: (a) If Finding Resolved.

If the In order to ensure In ensure that plant-specific plant-specific data collected in modification involves involves new new the collection data window is poolable poolable and applicable applicable equipment or a practice to the current plant, plant plant modifications (both where generic parameter hardware and procedural) implemented during this estimates are available, time period should be reviewed for potential potential impact USE the generic parameter USE for on this failure data. This review should be estimates updated updated with documented and the use of plant-specific data should plant-specific data data as itit be limited, limited, as appropriate.

becomes becomes available available for significant basic events; or (b) IfIf the modification (b) modification isis unique to the extent that generic generic parameter estimates estimates are not are not available available and only limited_experience_is limited experience is Page 20 of 55

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

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

F&O# Finding or ASME Reg.Reg. Guide 1.200 (Supporting Finding Description (summary discussion)

Finding Disposition Suggestion Category II Text Requirement) following the available follOwing ANALYZE 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.

HR-Al -01 HR-A1-01 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 (HPSI) 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 P&lDs were examined to identify and define descriptions and simplified P&IDs 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. these. susceptible to Type A (pre-initiator) events were screened from further status.

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.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 g Resolved.

Finding Resolved process identified for screening pre-initiator human failure events should be sufficient to identify most pre- pre accident HRAs.

HR-A2-Ol HR-A2-01 Finding IDENTIFY, through a review Miscalibration events for the containment pressure Miscalibration 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 potential miscalibration those calibration activities This is similar to the F&O for the restoration events events, including those for the HPSI, HPSI, LPSI, and containment spray system performed incorrectly that if performed except for the calibration events. The miscalibration miscalibration described inin this finding. The initial 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 automatic initiation of on the automatic 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, identify events. include all systems modeled in the PRA, which are listed in the Palisades System Notebooks. Once the initial initial systems list was assembled assembled, the Miscalibration of containment pressure signals Signals would system descriptions and system and simplified simplified P&IDs were examined to identify and impact auto auto start of HPSI, HPSI,LPSI, LPSI, and Containment define the Train/Function/Channel (TFC) (TFC) for the the system.

system. Those TFCs TFCs not Spray System (CSS). (CSS). ItIt might also also impact impact auto start of susceptible to Type Type A (pre-initiator) events were (pre-initiator) events were screened from further containment unit coolers and CIS coolers and CIS signals.

review. For For each of the unscreened unscreened TFCs TFCs identified, identified, a scoping event event was added to the PRA model. The scoping values were then used to Page 21 of 55

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

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

irting Finding or ASME Reg. Guide 1.200 (Supporting Finding Description (summary discussion) Disposition Requirement) Suggestion Category II Text Requirement)

Review each system for possible pre-accident Review 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 governing the maintenance, surveillance and testing of the procedures goveming 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 Notebook Volume 2 NB-PSA-HR Volume 2, "Palisades Palisades Evaluation (8).

Pre-Initiator Human Error Evaluation" [8].

Finding Resolved.

HR-C1-01 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 E.2-1A identified in Tables E.2-1 E.2-1B 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 designator 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, function, system train, or descriptions and simplified P&IDs were examined to identify and define Inconsistencies between the documentation and the component affected.) Train/Function/Channel (TFC) for the system. Those TFCs not the Train/FunctioniChannel model make reviews difficult and might lead to susceptible to Type A (pre-initiator) events were screened from further additional questions on modelmode adequacy.

a equacy.

NB-PSA-HR Volume review (this process is documented in Table 2.2-1 of NB-PSA-HR Update the HRA evaluations and the HRA document [8]). For each of the unscreened TFCs identified, a scoping event was II [8)).

to match the BEs 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- (NB PSA-HR Volume II (8)).[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 HFEs train/channel/function level of each system. The revised methodology and trainlchanneVfunction unscreened activity, result identified in F&Os against HR-A identified HR-Al1 and HR-A2, and no new pre-initiator HEPs are discussed in the HRA Notebook new Notebook Volume II [8).

[8].

from failure to restore (a) evidence of a review of plant specific specific mispositioning Those pre-initiators specifically identified Those identified during during the review and several equipment to the desired desired or miscalibration or events, credit cannot be given for miscalibration events, others were assessed using using the revised methodology and, and, as as necessary, standby oror operational collection of plant-specific plant-specifiC or generic generic operating operating were were added to the model.

added to model. A review of of plant history history was conducted for status (b) initiation status (b) initiation signal or or experience, experience. plant plant specific operating operating experience.

experience. The result of of the review was was that that point for equipment start-set point while there were instances noted noted of conditions that would be considered No_review_of_plant_misposition_or_miscalibration_and No reviELw_ of plant misp~ition or miscalibration ancj Page 22 of 55

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CFR SO.SSa(a)(3)(i)

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

F&O#

Finding or Finding or ASME Reg.

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

(summary discussion) Disposition Disposition Suggestion Suggestion Category IIII Text Category Text Requirement)

Requirement) up or up or realignment realignment (c) (c) missing events missing events generally generally included included for for standby standby pre-initiators, the pre-initiators, the examples examples noted noted were were either already covered either already covered by by aa automatic realignment automatic realignment or or components and components and instrumentation instrumentation as as discussed discussed inin HR-HR- pre-initiator event pre-initiator event identified identified during during the the implementation implementation ofof the the revised revised power ADD power ADD failure failure modes modes Al and A1 and HR-A2.

HR-A2. methodology or methodology or were were related related to to equipment equipment notnot credited credited in in the PRA.

PRA. The identified during identified during the the plant operating experience plant experience review review is is documented documented in in HRA HRA notebook notebook collection of of plant-specific plant-specific or or Perform aa systematic review Perform review of of HFEs.

HFEs. Consider aa collection volume 11 [251.

volume [25].

Report review Condition Report review of mispositioned or applicable generic applicable generic operating miscalibrated events miscalibrated events to determine ifif any any trends trends Finding Resolved.

Finding Resolved.

experience that experience that leave leave equipment unavailable associated with the associated with the pre-accident events could pre-accident events could impact equipment unavailable for for response in in accident accident the HRA values.

the response sequences.

sequences. I HR-D4-0l HR-D4-01 Suggestion Suggestion When taking into When into account Recovery factors are credited in the detailed The HFE summary table, Table 4-1 of NB-PSA-HR Vol. 2 [81, [8], was revised self-recovery or recovery self-recovery recovery evaluations of pre-initiator human failure events to provide a listing of the recovery factor ASEP cases applied to from other from other crew crew members in (HFEs). However, the write-up is not clear which of pre-initiators HFEs for which detailed analysis was performed.

estimating HEPs estimating HEPs for for specific specific the ASEP cases that the evaluation represents. The HFEs, USEUSE pre-initiator pre-initiator Suggestion Resolved HFEs, HRA Calculator shows the associated ASEP case in recovery factors in a manner recovery manner the detailed evaluation. However, the information is consistent with selected consistent not presented in the document which makes review methodology. If recovery of somewhat more difficult.

difficult.

pre-initiator errors pre-initiator errors is is credited credited ESTABLISH the (a) ESTABLISH the This is a documentation issue only.

(a) maximum credit credit that can be Update the HRA calculator write-up to be clearer as to given for given multiple recovery for multiple the ASEP case used for the detailed evaluations.

opportunities (b) USE the opportunities following information to following information assess the potential potential for recovery of pre-initiator: (1) recovery post-maintenance or post-maintenance or post-calibration tests required and performed by procedure and performed procedure (2) independent (2) independent verification, using a written using written check-off check-off list, that verifies verifies component component status following maintenance/testing (3) a separate check separate check of of component component statusstatus made made at a later time, later time, using using a written check-off check-off list, by the original list, by original performer performer (4) (4) work shift or work shift or daily checks of daily checks of component component status, using aa written status, using written ~~---

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

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

I Finding Disposition Suggestion Category II Text Requirement) check-off list.

check*off HR-E1-01 Suggestion When identifying the key There does not appear to be a systematic review of routinely been reviewed Procedures have routinely reviewed by system analysts and HRA human response actions relevant guidance in the procedures and other relevant 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 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' 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.

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 following in the event response.

procedure(s) the operator would be follOwing 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 y stem is obtained The procedural guidance is documented in both the PIOAQ form and in system 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 Suggestion When identifying the key In Table 2-1 HRA HRA Summary Table, for Type C: The guidance provided in in the HRA notebooks indicates indicates that the penalty of human response actions Procedural Actions During Course of Accident, it 1.0 human error probability (HEP) for these actions is overly a 1.0 REVIEW: (a) the plant- the evaluation of actions with no states that Uthe 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 1.0 for non- that since the guidance hasn't hasnt actually been applied to any current HEP other relevant procedures proceduralized actions; however, this is judged to be values, the guidance guidance itself should be removed to avoid confusion. The (e.g., AOPs, annunciator annunciator overly conservative when the time available available for action action purpose ofof this wording in in the notebook is to provide guidance guidance for a variety response procedures) in the response and the TSC becomes available.

is long and available.* of of circumstances that maymayor or may not not be applicable at at any given time. As context of the accident context accident the guidance guidance was not not implemented at at the time of the PEER PEER review does does scenarios (b) system However, itit doesn doesn'tt appear appear that any any operator actions actions not mean itit could not not not be implemented later.

operation used this.

operation such such that an -

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

F&O# Finding or Finding ASME Reg. Guide 1.200 (Supporting Finding Description (summary discussion) Disposition Suggestion category II Text Category Requirement) understanding of how the This is primarily a documentation issue. Statement Suggestion Resolved system(s) functions and the contusion if not used.

causes confusion human interfaces with the system is obtained Although non-proceduralized operator actions don't 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 (Le., (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 & III. Ill. 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 Ill.

required by Cat II & III. (10 per crew) for review. Their reviews included ensuring indications,

(-10 with plant observations and training procedures. procedure selection and use, and activity performance man-power and Document the talk through performed with training 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 1 [25].

Finding Resolved.

HR-G6-01 Finding CHECK the consistency consistency ofof HRA Procedure Procedure 5.3.2.12 states: The"The consistency consistency of A comparison of of the human error error probabilities (HEPs) developed for each HEP the post-initiator HEP Error Probabilities resulting post-initiator Human Error human failure event (HFE) in in the PLP internal events PRA model shows quantifications. REVIEW REVIEW the (HEP5) should be checked: (a)

(HEPs) (a) REVIEW the Human that the values of the HEPs HEPs are internally consistent relative to each other, HFEs and their final HEPs Failure Events (HFEs)

(HFE5) and their final HEPs relative to and generally follow a trend of lower HEPs HEPs being being associated associated with lower relative to each other to each other to check their reasonableness given the stress stress levels levels (which in turn tum may be associated associated with more more time available to check their reasonableness scenario context, plant history, procedures, take action).

action). Exceptions to the general trend are present, present, and can be be given the scenario context, given operational operational practices, and experience. (b) One explained explained through detailed examinations of the contributions to the HEP plant history, procedures, approach for checking the consistency consistency of HEPHEP (e.g., number of procedure procedure steps, steps, time available available to perform perform the steps, operational practices, practices, and and quantifications quantifications is to sort is to by increasing sort by increasing or decreasing probability of successfully successfully recovering from errors errors occurring during during experience.

experience. HEP HEP values values and then performing the comparison.

comparison.* InIn completion of the procedure, procedure, etc.). This review is is documented documented inin Section addition, HRA HRA Notebook Notebook Section 4.0 states: After Section 4.0 "After the 4.0 of NB-PSA-HR Volume 4.0 of Volume 11 [25].

[25].

individual results were were obtained, the final HEPsHEPs were were assessed for appropriateness Finding Finding Resolved.

Resolved.

appropriateness and and consistency within - --

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

F&O#

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion) Disposition Suggestion Category II Text Requirement) the PLP HRA. Human action element such as time actions diagnosis and/or frame and complexity of the action's execution were considered. When available, HEP results for similar actions at other PWRs were used reference. However, no as further points of reference."

documentation of the review was found.

Consistency check is required for Capability Category Ill. Document the consistency check that was I, II, and III.

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 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 au- QU-Cl, ASSESS the degree of human reliability analysis of the PSA developed human error probabilities C1, After the HRA is complete, redo and document the (HEP5) as though they were independent of one another. It is known that (HEPs) dependence, and calculate evaluation, 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 dependence. ACCOUNT for the core damage frequency may be higher than quantified in the accident sequence analysis. This analysis will evaluate the post-initiator the influence of success or 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 significant. The most risk significant human error damage frequency is Significant.

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 available to perform the 1. Run the base model with the post-initiator post-initiator action failure event actions (b) factors that could probabilities set to 1.0.

1.0.

lead to dependence (e.g., 2. Identify the multiple human action combinations that appear appear in in the cut common instrumentation, sets.

common procedures, increased stress, etc.) (c) 3. Identify the risk significant combinations assuming complete 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 Incorporate the dependent dependent combinations in in the fault trees of of the PSA.

PSA.

To address the human actionaction dependency issue issue with respect to CDF, CDF, Palisades_developed_a_systematic_approach_that_investigates_a_sufficient Palisades developed a systematic approach that investigates a sufficient Page 26 of 55

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

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

F&O #

I F&O# Finding or 1.200 (Supporting Finding ASME Reg. Guide 1.200 Finding Description (summary discussion) Disposition Suggestion Category IIII Text Requirement) 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. finding is internal events PRA.

still open for the full power intemal Finding Open for full power intemal internal events HEPs.

HR-Hi -01 HR-H1-01 Suggestion INCLUDE operator recovery aESS-XVOT-SWS-ESS 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 significant sequences to determine if review of Significant procedures and cues that occur duringdunng the sequence. The review of significant accident Significant 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 SA.

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 SR.

this SA.

Consider adding documentation of this review to the next major PRA model update.

HR-l3-0i HR-13-01 Finding DOCUMENT the sources of There are only two assumptions in the entire HRA Table 1.6.1 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 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 QU-El and au-QU- defaults and methodology requirements are not listed assumptions are categorized into fire related and general HRAHRA methods E2) associated with the E2) as assumptions 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 ofof model model uncertainty.

uncertainty.

Review Review the HRA HRA for additional additional imbedded imbedded assumptions and_use_the_updated_list_for_potential_model and use the updated list for potential model -

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ENO, PALISADES ENO, PALISADES NUCLEAR NUCLEAR POWER POWER PLANT PLANT 10 CFR 10 CFR SO.SSa RELIEF REQUEST 50.55a RELIEF REQUEST NUMBER NUMBER RR RR 4-204-20 PROPOSED ALTERNATIVE PROPOSED ALTERNATIVE IN IN ACCORDANCE ACCORDANCE WITH WITH 10 10 CFR CFR SO.SSa(a)(3)(i) 50.55a(a)(3)(i)

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

F&O#

Finding or Finding or ASME Reg.

ASME Reg. Guide Guide 1.200 1.200 (Supporting (Supporting Finding Description Finding Description (summary (summary discussion) discussion) Disposition Disposition Suggestion Suggestion Category IIII Text Category Text Requirement)

Requirement) uncertainties.

uncertainties.

IE-A6-01 IE-A6-01 Finding Finding When performing When performing the the Although itit appears that anan evaluation evaluation of CCFs was was AA detailed evaluation evaluation to to address this this finding was was completed completed in in Attachment systematic evaluation systematic performed since IE_LOY1Q-Y20 performed IE_LOY1 0-Y20 and Reference [9]. In summary, the evaluation states:

3 of Reference required in required in IE-AS, IE-A5, INCLUDE INCLUDE lE_LO-ALL4PREFAC, etc, IE_LO.,ALL4PREFAC, etc, were identified; however, however, events resulting initiating events resulting documentation of of the systematic systematic evaluation for the AA process process to to ensure ensure all possible common cause combinations and and routine routine initiating documentation multiple failures, from multiple failures, ifif the the other support system system CCF CCF events events was and non-routine system alignments and alignments is theoretically achievable, achievable, but time time from elimination of other equipment failures failures result result not provided.

provided, consuming and inconsistent consuming inconsistent with risk-informed approaches as as utilized utilized in equipment not common cause, and from a common and PRAs. However, aa process PRAs. process to ensure ensure that combinations of failure events from from routine Provide documentation of the evaluation of electrical and system configurations that have occurred or could reasonably occur is from routine system alignments, events, 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 speCifications.

combinations if in fact such combinations are necessary to result in a plant trip.

trip.

Finding Resolved.

IE-A6-02 IE-A6-02 Finding Finding When performing When performing the the Event trees for common cause failures (e.g., Loss of A detailed evaluation to address this finding was completed in Attachment systematic evaluation systematic evaluation Preferred AC Bus Y20, Y30, and Y Y40)

40) are included in 3 of Reference [9].

required in required in IE-AS, IE-A5, INCLUDE the SAPHIRE program, but no documentation events resulting initiating events In summary, the evaluation states:

initiating associated with these event trees has been found.

from multiple from multiple failures, failures, ifif the the Note, the FMEA discussion provided in the Initiating A process to ensure all possible common cause combinations and routine equipment failures equipment failures result Event notebook, does not specifically discuss the and non-routine system alignments is theoretically achievable, but time common cause, and from a common CCF initiators, nor nor does it identify identify the buses buses as consuming and inconsistent with risk-informed approaches as utilized in from from routine routine system system necessarily resulting inin a reactor trip. No discussion of However, a process to ensure that combinations of failure events PRAs. However, alignments, alignments. non-routine system alignments has been found. and system configurations that have occurred or could reasonably occur is in place already through the current Palisades approach of considering The NRC's NRCs clarification for this element requires consideration, and documentation of Initiating genenc data, initiating event categonzation, plant and generic categorization, and technical and documentation Initiating events resulting from common specifications. This specifications. This process addresses reasonable common cause common cause or from both both routine combinations ifif in in fact such combinations are are necessary to result in in a plant and non-routine and non-routine system system alignments.

alignments.

trip.

systematic approach to ensure A systematic ensure all possible possible common cause combinations and routine and non-routine Recognize that given the plants asymmetries the CCF grouping is system alignments straightforward.

straightforward.

alignments needs needs toto be developed, developed, and and documented.

documented. These results results indicate indicate that the impact impact of CCF (e.g.,

(e.g., the preferred acac buses) buses) is is minimal when considering minimal when conSidering random failures. The The PRA PRA isis satisfactory satisfactory to to achieve at at least Category II compliance least aa Category compliance with the ASME Standards Standards IE IE A5 AS and and A6.

Finding Finding Resolved.

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

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

F&O# Finding or Finding ASME Reg. Guide 1.200 (Supporting Finding Description (summary discussion) Disposition Suggestion Category II Text Requirement)

IE-A8-01 Finding INTERVIEW plant personal INTERVIEW No meeting minutes or documentation of reviews NB-PSA-IE [10] was revised to document interviews and Section 2.2 of NB-PSA-IE (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 calculation owners. Interviews with System Engineers were performed by Lack of documentation of reviews performed by overlooked, overlooked. the PRA the PRA personnel and documented in Attachment 5 of all PRA system Licensed operators, system engineers and Licensed 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 initiating events or the Special Initiators as *Special added to Section 2.2.6 of the initiating events notebook NB-PSA-IE [1 OJ to

[10]

experience for initiating potential for such events (e.g., precursors) was determine if they are potential precursor events.

determine events. Component failures were event precursors, for performed during the PRA teams'teams review of the obtained from Attachment 3 of the data notebook NB-PSA-DA [5J [5] and identifying additional Maintenance Rule (MR) database and Maintenance individually evaluated as to their potential as a precursor event. No new individually initiating events. For Work Orders (MWO) in support of the data effort.* 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.

provided, appropriately modeled in the PRA.

structure clogging might Provide documentation to show the evaluation Finding Resolved.

indicate that loss of intake performed.

structures should be 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 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.

groupings.

The criteria added consider:

considered similar inin 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 1.

criteria, timing, and the actions, sufficiently different to be considered in different effect on the operability and groups. 2. the initiating event disables instrumentation which is required for performance of operators Explicitly include consideration of operator action successful operator action, oror and relevant mitigating systems; or (b) events cancan timing in defining the initiator groups.

groups. initiating event

3. the initiating event changes the likelihood likelihood of successful operator be subsumed into a group group performance by some other mechanism mechanism and bounded by by the worst worst case impacts within the Suggestion Resolved 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 #

F&O# Finding or Finding ASME Reg. Guide 1.200 1.200 (Supporting Finding Description (summary discussion) Disposition Disposition Suggestion Category II Text Requirement) new group. 00 "new" 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 IE-94-01 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 nsk of pressurized thermal shock. These efforts are to re-evaluate the risk categories those categories Thermal Shock (PTS) evaluation. However, an NUREG-1 806 and NUREG-1874. The analyses made use summarized in NUREG-1806 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 SA. 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 (the generic frequency) frequency*) to model Excessive have more severe LOCAl Vessel Rupture in lieu of the latest plant specific state-of-radionuclide release LOCA/Vessel Rupture should be included Excessive LOCANessel knowledge based on the joint RES/Industry 50.61 initiative is not in the model as leading directly to core damage.

potential (e.g., LERF). This warranted. Note that the dominate sequence was a non-mechanistic Palisades can use the generic frequency or they can includes such initiators as scenario that assumed the pressurizer pressunzer safeties failed open for a period of use the frequency from their Pressurized Thermal excessive LOCA, interfacing time and subsequently reclosed. The next set of dominant sequences did Shock Analysis.

systems LOCA, steam not include a pressure component. Refer to NB-PSA-IE, Palisades "Palisades generator tube ruptures, and Probabilistic Safety Assessment Initiating Event Notebook",

Probabilistic Notebook, [10].

unisolated breaks outside containment. NB-PSA-IE dedicates 4 pages addressing addreSSing Pressurized Thermal Shock.

Palisades was one of three pilot plants evaluated in the NRC Palisades NRC initiative to re-evaluate the risk ofof pressurized thermal shock. The analyses made use use of three Palisades specific specific analytical analytical models models (PRA, RELAP and and FAVOR) that together, allowed the estimate of the yearly through-wall crack frequency (TWCF)

(lWCF) in in the reactor pressure vessel (RPV).

(RPV).

Suggestion Resolved IE-Ci-Ol IE-C1-01 Suggestion CALCULATE the initiating This is is in Reference to Section 3.1 of the Initiating Initiating Subsequent to performance of the peer review, NUREG/CR-7037, NUREG/CR-7037, event frequency frequency accounting Notebook (NB-PSA-lE):

Events Notebook (NB-PSA-IE): TheaThe thermal capacity Industry "Industry Performance Performance of Relief Relief Valves at U.S. Commercial Nuclear Nuclear Power for relevant generic generic and and of the steam generators generators at Palisades is such such that a Plants through 2007, 2007," was published. This document document provides provides an an update update plant-specific plant-specific data data unless unless it demand on on the the PORVs PORVs or or pressurizer SRVs SRVs isis not not of the industry data related to safety valves and industry data and low low capacity capacity relief valves, is justified that there are are expected following a reactor trip. This has been as well as illustrating an approach approach for modeling pressurizer safety valves, adequate plant-specific datadata validated per review of of past thermal hydraulic hydraulic main main steam safety valves, atmospheric dump steam safety dump valves valves and PCS PCS PORVs in in to characterize the analyses (Final Safety Analysis (Final Safety Analysis Report (FSAR) PWR PRAs.

parameter parameter value and and its Chapter Chapter 14).

14). In In addition, in in the 30 plus plus years ofof -

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

F&O# Finding or Finding ASME Reg. Guide 1.200 (Supporting Finding Description (summary discussion) Disposition Requirement) Suggestion Category IIII 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),

events). of all PWRs. Only inadvertent or premature operation event, but is considered as a consequential or spurious spunous event.

of these valves can lead to loss of coolant type Suggestion Resolved conditions. Given a demand and subsequent failure of SRVs, the consequences of a small the pressurizer SRV's, 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 317/94-007)]

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-IE-LOCA PZRSRV is consistent with the definition and events NUREG/CR-6928s for IE-SORV used to calculate NUREGlCR-6928's lE-SORV NUREG/CR-i6928 lE-SORV (PWR). The events in NUREGlCR...,6928IE-SORV were used directly 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 IE-C2-01 Finding Finding When When using using plant-specific plant-specific Justification Justification for the exclusion of data before before January Added additional justification justification for the the exclusion of data prior to to January January data, USE data, USE the most most recent 2003 2003 used used to identify identify plant-specific initiating initiating events events 2003 2003 toto Section 4.1 of the initiating initiating events notebook NB-PSA-IE NB-PSA-IE [10].

applicable data to quantify was not provided. Justification Justification is based on improved plant availability from January 2003- 2003 the initiating initiating event 2009 relative to the previous site specific initiating event data from Justification Justification for the exclusion of data before January frequencies. JUSTIFY January 1994 - December 2002. Improvements in plant availability were January 1994 2003 2003 used to identify identify plant-specific initiating events excluded data that is not demonstrated graphically in Figure 4.1. Plant availability has was not provided.

provided, considered to be either either demonstrably improved after January 2003 due improved operating and January 2003 recent or applicable (e.g., Provide the requested justification. maintenance practices.

provide provide evidence via design Finding Resolved.

or operational change that or the data data are are no no longer longer applicable.)

lE-C6-01 IE-C6-01 Finding Finding USE as USE as screening criteria nono In relation to lE-C6, IE-C6, Operator actions are apparently The basis excluding control room HVAC basis for excluding HVAC from the full power power internal internal higher higher than the following credited credited for the exclusion of of some events (e.g., events model model was strengthened strengthened to include include other other aspects aspects in in addition 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)

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

F&O# Finding or ASME Reg. Guide 1.200 (Supporting (Supporting Description (summary discussion)

Finding Description Disposition Suggestion Category II Text Requirement) characteristics (or more CRHVAC refer to earlier HVAC comments) without NB-PSA 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 components and concluded that a loss of HVAC would not result in a If component/system failures lead to an initiating groups from further significant increase in the failure probability of the RPS.

event but are screened from further analysis by evaluation: (a) the frequency crediting operator actions or equipment/systems to In addition, a comparison of sensitivity analyses performed based on 14 of the event is less than 11 E-E

(/ry), and avert the transient, then quantify the total initiating owners group sites that modeled the contribution to CDF due to loss of owner's 7 per reactor year (fry),

event frequency considering these events and apply control room HVAC. The sensitivity studies found that the average CDFfyr CDF/yr the event does not involve criteria of IE-C6 to determine if screening criteria is was 1.61 E-07 with a median of 1.31 E-07fyr.

E-07/yr. Given Palisades core either an ISLOCA, met. damage frequency is on the order of E-05, the change in CDF due to loss containment bypass, or of control room HVAC would less than 11%.  %.

reactor pressure vessel Apply IE-C6 screening criteria and document as rupture (b) the frequency of appropriate. With respect to cable spreading room cooling. An analysis of the cable I the event is less than 11 E- E I

spreading room heat-up following a loss of ventilation was developed 6/ry, and core damage could 6fry, EA-PSA-GOTHIC using the GOTHIC software code and documented in EA-PSA-GOTHIC-not occur unless at least two 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 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 spreading and control room areas is not necessary necessary to be explicitly event conditions, with a high degree modeled and the bases for these conclusions do not require operator modeled degree of certainty (based on supporting calculations), action action to mitigate elevated temperatures. However, ventilation is is are detected and corrected considered for purposes of fire modeling modeling in these areas.

before normal plant Finding Resolved.

Finding operation is curtailed (either (either administratively or automatically). IfIf either criterion (a) or or (b) above is used, used, then CONFIRM that the value specified in the criterion meets meets the applicable requirements in applicable in Data Analysis AnalysiS (2-2.6) and Level 11 Quantification (2-2.7).

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

F&O# or Finding or ASME Reg. Guide 1.200 (Supporting Finding Description (summary discussion) Disposition Suggestion Category II Text Requirement)

IE-C12-01 Suggestion COMPARE results and NB-PSA-lErl describes the quantification Section 5 of NB-PSA-IEr1 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 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.

the results.

check of the 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 NUREGlCR-5750 concem 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 internal events analysis) were nearly the same as that the current intemal vintage, 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.

NUREG-1 829.

Suggestion resolved.

lE-C14-01 IE-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 Le., CK-ES31 01, CK-ES3116, i.e., CK-ES3101, CK-E531 16, CK-ES3131and 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 potential initiators because of a pair of valves as potential ISLOCA through 5 containment penetrations with 17 17 potential 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 used 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 ISLOCAISLOCA analysis (d) the capability of documentation to clearly demonstrate the fact that documentation secondary system piping (e) isolation these tests were sequenced sequenced such such that the valves isolation capabilities given were were confirmed closed closed on a quarterly basis. This was was high high flow/differential not immediately immediately apparent apparent in in the documentation.

pressure pressure conditions conditions that that might_exist_following_breach might exist following breach Page 33 of 55 of

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

F&O#

Finding or Finding or ASME Reg.

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

(summary discussion) Disposition Disposition Suggestion Suggestion Category IIII Text Category Text Requirement)

Requirement) of the the secondary secondary system system LE-C9-01 LE-C9-01 Suggestion Suggestion JUSTIFY any credit given JUSTIFY any given No credit No credit is is taken for equipment equipment operability operability or No credit is No is taken taken for for equipment equipment operability or or operator operator actions actions inin adverse adverse equipment survivability or for equipment or actions for operator actions for adverse adverse environment or environments or after after containment containment failure.

failure. Palisades Palisades reviewed reviewed the the LERF LERF human actions under human failure. In containment failure. In Section 6.2.4 of of the Level Level 2 results for for opportunities opportunities to take such credit (as documented documented in in Section adverse environments. report, Palisades stated stated that they had reviewed the 6.2.4 of the Level 2 Notebook) and and justified the lack of credit.

credit.

results for results for cases where credit for equipment equipment or HAAs HRAs harsh environment or after containment Based on on way the standard is is written, the only only way to eamearn aa CC-II CC-Il during harsh containment failure categorization is categorization is to credit credit equipment equipment operation in adverse environment environment (for(for be applicable but might be did not justify equipment but did LE-C9 and C-10) C-i 0) and after containment failure (for (for LE-C11 LE-Ci 1 and C12).Cl 2).

survivability in either of these conditions based based onon the 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.

This clearly meets the requirements for Capability The MAAP program was utilized in calculation PLP0247-07-0004.01R2 PLP0247-07-0004.0i R2

[14] to determine the bounding best-estimate containment environmental I

Category II conditions postulated to be encountered by equipment located in To move up to Capability Category 11/111, Il/Ill, 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 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-455-00i-DC1 CALC-455-001-DC1 [15] evaluates the survivability of modeled in the PRA under the environmental equipment modeled environmental conditions determined determined in the MAAP analyses. analyses. This analysis analysis utilized temperature temperature profiles from the MAAP program to demonstrate that all profiles all credited PRA PAA equipment located located in in containment can can survive survive the limiting limiting containment conditions produced by by MSLB, LOCA, and OTC scenarios in which only only a single containment air cooler or a single Single containment spray pump and header are available.

available. A A further detailed detailed summary is is provided in in Attachment 88 of NB-PSA-ETSC Attachment NB-PSA-ETSC [11]. [11].

In summary, In summary, itit isis considered that that supporting requirements LE-C9 LE-C9 andand LE-ClO LE-C10 meetmeet the CC-Il CC-II requirements as as the above above noted noted engineering engineering evaluations provide evaluations provide the justification.

justification.

Palisades Palisades does does notnot take take credit credit for for continued continued operation operation of of equipment equipment or or operator action operator action after after containment containment failure.

failure. Therefore, Therefore, by by definition, definition, aa CC-I CC-I Page 34 Page 34 ofof 55 55

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

F&O#

F d Finding or ASME Reg.

R G *d e 1.200 Guide 1 200 (Supporting Suggestkn (summary discussion)

Finding Description (summary Disposition Suggestion Category II Text Requirement)

LE-Ci 1 and C12 classification of supporting requirements LE-C11 Cl 2 is considered appropriate.

Finding Resolved LE-G5-01 LE-GS-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.

applications, 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 11 analyses, but the level Finding Resolved 1 discussion does not explicitly cover LERF so their 1

analysis does not comply with the SA. SR.

Palisades should develop such a discussion similar to that developed for the level 11 analyses or revise the level 1 1 discussion to include LERF.

MU-B2-01 Finding changes that would impact Changes 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 "PSAR3 Release 2b". 2b.

should be prioritized to internal flooding analysis. The current release for the intemal 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 1.200, NFPA-805 issues from multiple spurious significant changes are Significant 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 1.200 as well as was described during the peer review 10/26/2009 Monday moming morning practical. NFPA-805. PSAR3 Release 2b is changes to address NFPA-80S. 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 lE frequencies etc.,

updating to addressing extensive flow diversion scenarios, to adaptation of updating the simplified Westinghouse LERF model, to incorporation of a new comprehensive common cause model that employed employed the latest data, etc.

PSAR3 PSAR3 Release 2b differed from PSAR3 Release 2a due to inclusion of new IE, new IE, HRA data etc.

With exception of the last significant 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 dateddated 6-30-2006.

6-30-2006. Both PSAR3 Release 2# series included the the finalized GSI-191 GSI-191 modifications.

modifications. These These GSI-191 GSI-191 physical physical modifications were were completed completed in in the the spring spring of of 2009 during the scheduled 2009 during scheduled REFOUT.

REFOUT. These These modifications included an extensive re-analysis re-analysis of of the the passive_screen_design_due_to_the_reconciled_chemical-effect_tests_that passive screen design due to the reconciled chemical-effect tests that -

Page 35 of 55

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

p#orting or Finding or ASMEReg.Gtde 1.200 Finding ASME Reg. Guide 1.200 Finding Description (Supporting Finding Description (summary (summary discussion) discussion) Disposition Disposition Requirement) Suggestion Category II Text Requirement) occurred in occurred in the the fall of of 2008.

2008.

Given the PRA Given PRA teams design design basis basis responsibilities, support of responsibilities, support of this this initiative was initiative was very involved involved and included included support of anan on site NRC onsite NRC inspection that was inspection was required required for startup approval.

approval. The The GSI-191 mod mod was most significant plant change that had the most had occurred since release release of the dated 6/30/2006 (PSAR2c).

analysis-of-record dated current analysis-of-record Release 2c has Release has since been completed and and the Palisades Palisades QA process applied in establishing the anew a applied new analysis-of-record analysis-of-record termed PSAR3 to support the NFPA-805 initiative. This model support model version is used as the base 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.

GSI-1 91 modification addressed the uncertainty associated with The GSI-191 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 MU-B3-01 Suggestion Suggestion PRA changes PRA changes shall 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 performed 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 upgrades against the ASME standard.

Supporting Requirements.

The configuration control document document does does not Suggestion Resolved specifically indicate that updates are to be done in accordance with corresponding SRs from the standard, but it is assumed assumed that the definition of appropriately implemented

-appropriately implemented" includes includes such such as review because the associated system, system, IE, IE, or or other notebooks that would be notebooks be updated updated all currently have have a section for self-assessment against the standard.

Add Add a sentence to the the configuration configuration control control document document to clarify that to clarify appropriately implemented that -appropriately implemented" means means conformance to the the standard standard supporting supporting requirements.

requirements.

MU-B4-01 MU-84-01 Finding Finding PRA Upgrades PRA Upgrades shall shall receive receive The Configuration The Configuration Control Control Notebook Notebook specifies specifies the the Section Section 3.33.3 of the configuration of the configuration control control notebook notebook NB-PSA-CC NB-PSA-CC [17] [17] has has a peer peer review review (in (in difference differen~ between bet\Y~n and and update update and and an an upgrade upgrade but but been revised to been to include include aa requirement requirement for for aa peer peer review review against against the ASMEASME Page Page 36 36 of of 5555

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

F&O# Finding or Finding ASME Reg. Guide 1.200 (Supporting Finding Description (summary discussion) Disposition Disposition Suggestion Category IIII Text Requirement) accordance with the performance of a peer does not specifically require perfonnance 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 PRA Upgrades, but the Configuration Control of the PRA that have been Notebook does not specifically call for one following upgraded. Refer to Section an Upgrade.

2 of the ASME PRA 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.

update.

MU-Di-Ol MU-D1-01 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.

risk-infonned upgrades against previous applications and analyses.

changes on previously Review of previous RI applications is not called out in Finding Resolved implemented risk-infonned risk-informed the Configuration Control Notebook.

decisions that have used the PRA AND PRA AND that affect affect the requirement for reviewing the previous RI Add requirement safe operation of safe operation of the the plant.

plant. against the new PRA results to see if applications against they impact thethe results of the previous previous work.

QU-Al-Ol QU-A1-01 Suggestion Suggestion INTEGRATE the INTEGRATE the accident accident Figure 5-1 Figure 5-1 of of the Quantification Report provides a To be implemented in a future notebook revision. This is a documentation sequences, system sequences, models, system models, chart on the process of integrating the small flow chart issue only. No impact to the interim FPIE application for RI-ISI screening.

data, and data, HRA in and HRA in the models into the SAPHIRE code and the CAFTA models quantification Suggestion Open process for quantification process for additional APIs additional prepare the used to prepare APls used the SAPHIRE model each initiating each initiating event event group, group, for quantification (including for (including the the integration of CCF accounting accounting for for system system rules, etc.) While this flow chart gives an trees, HRA rules, dependencies, dependencies, to arrive arrive at upper level explanation of the process, level explanation process, a more accident sequence accident sequence detailed flow chart would be useful in in ensuring a frequencies.

frequencies. consistent integration for personnel that do not consistent perform this task frequently.

perfonn The integration integration process is fairly complex and involves involves multiple codes multiple codes and tools. Missing Missing any step in this process could impact quantification.

Develop a more more detailed flow chart chart for those performing the quantification.

perfonning QU-A3-01 QU-A3-01 Finding Finding ESTIMATE ESTIMATE the mean CDF the mean CDF The mean ISLOCA The mean ISLOCA CDF frequency does does not not account A method of demonstrating demonstrating the effect effect of the state of of knowledge knowledge is is to accounting accounting for for the state-of-state-of- for the state-of-knowledge correlation (SOKC). Per Per perfonn a Monte perform Monte Carlo simulation for representative cases. Given the knowledge knowledge correlation SR QU-A3, the effect of the SOKC has been found to reference to ISLOCA frequency fr~quency in the ASME ASME Standard and the finding, Page 37 of 55

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Table A2.S-1, A2.5-1, March 2010 Full Power Internal Internal Events Peer Review Report Findings Findings and Resolutions F&O# Fmdingor 1.200 (Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion)

(Supporting Disposition Suggestion C Category II Text Requirement)

Requirement) between event between event probabilities probabilities be significant in cutsets contributing to ISLOCA two examples were selected using as input the failure rate and when significant when significant [Note

[Note (1(1)].

)]. frequency. Explicitly required in Note 11 of the SA.

frequency. SR. distributions from the PSAR3 model, namely: ECCS injection line check Update the ISLOCA frequencies with SOKC. valves FTRC and SOC 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 CAFT CAFTA A containing the following components and failure modes when generating results with follOwing point factors:

2 MOVs FTRC - SOKC factor =

- =3 (SDC suction) 3 (SOC 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 SDC lines dominate the ISLOCA results. A factor (SDC) or 4 (LPSI) is not a significant deviation, particularly for of 3 (SOC) applications where uncertainty analyses are perfonned performed as a part of the incorporating the suggested rules file into the evaluation. Because incorporating SAPH IRE model results in a negligible SAPHIRE negligible impact on overall core damage

( within the uncertainty frequency (within 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 incorporated ISLOCA events.

Finding Finding Resolved QU-B2-01 QU-B2-01 Finding Finding TRUNCATE TRUNCATE accident Palisades used a truncation level of 1 1 E-09 for An analysis to determine detennine the full power internal events model power intemal model truncation sequences sequences and and associated quantification quantification and and conducted conducted evaluation of limit will will be documented documented in Section Section 6.6 of PSA notebook NB-PSA-QU NB-PSA-QU after system system models models at at a convergence of of the the results down to a truncation level truncation level issuance of the complete complete model report, including including human human error error sufficiently low cutoff value of 11E-1 2. The truncation E-12. truncation should be be set set to 11E-l E...,111 dependency, dependency, and full cutset cutset review review of of all all event event trees.

trees.

that dependencies based on on the Palisades definition of of significant This approach approach will apply the ASME PRA PRA standard HLR-QU-B HLR-QU-B whichwhich associated with significant accident accident sequences.

sequences. states, states, convergence "convergence can be be considered sufficient whenwhen successive cutsets or or accident accident reductions in in truncation truncation value of one decade result in in decreasing decreasing changes sequences sequences are not not in in CDF CDF or or LERF, and the final change change is less less than than 5%

5% which indicate indicate that eliminated. NOTE:

NOTE: a truncation of of four four orders orders of of magnitude below below the the CDF CDF is is adequate adequate for for aa Truncation Truncation should be be high high quality quality PRA.

PRA".

carefully carefully assessed assessed in in cases where where cutsets cutsets are are merged merged to A A truncation truncation study study hashas been been completed completed for the flooding PRA as as create create aa solution solution (e.g.,

(e.g., documented documented in in EA-PSA-INTFLOOD-13-06 EA-PSA-INTFLOOD-13-06 Rev. Rev. 00 [43].

[43].

where where system system level level cutsets cutsets are are merged merged to to create create .

sequence sequence level level cutsets).

cutsets). Finding Finding OpenOpen Page Page 38 38 of of 55 55

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

F&O#

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

(summary discussion) Disposition Disposition Suggestion Suggestion Category IIII Text Category Text Requirement)

Requirement)

QU-C1-01 QU-C1-01 Finding Finding IDENTIFY cutsets with IDENTIFY with Conditional HEPs were developed Conditional HEPs developed byby Palisades Palisades for The complete The complete detailed detailed methodology methodology for evaluating evaluating human human error multiple HFEs multiple HFEs that that several HFEs several HFEs and incorporated incorporated inin the fault tree dependency is is described in in HRA HRA Notebook Notebook NB-PSA-HR NB-PSA-HR Volume Volume 11 [25].

[25].

potentially impact potentially impact significant models. Some models. Some accident accident sequences sequences revealed revealed HFE HFE The dependency The dependency analysis for fire relatedrelated HEPs HEPs are are documented documented in in section section I accident sequencesl accident sequences/ cutsets combinations for which combinations which dependency dependency between the between the 6.4 of notebook.

of the notebook.

by requantifying the PRA by PRA HFE5 has HFEs has not been assessed and documented. The general steps used in this analysis are as follows:

model with HEPHEP values setset sufficiently 1. Run the

1. the base base model with the post-initiator post-initiator action action failure event event to values that are sufficiently cutsets are not probabilities probabilities set set to 1.0.

1.0.

high that the cutsets not While the Palisades modelmodel has has been been quantified and truncated. The final cut sets for accident sequences have been been identified, 2. Identify the multiple multiple human human action combinations that appear appear in 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 analysis on the risk Significant 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 EA-PSA-INTFLOOD-13-06 Rev. 0 [43]. This finding is still open for the full power internal intemal events PRA.

Finding Open for full power internal intemal events HEPs.

QU-D1-01 QU-D1-01 Finding Finding REVIEW a sample of the The final model review has not been completed and has not The documentation documentation of the final model is complete with exception of the significant accident accident documented.

documented. The final review of accident sequence sequence final results cutset cutset review for the the remaining full power power internal intemal events sequences/cutsets sufficient sufficient results has has not not been been completed and and documented so initiators with human human error dependency incorporated. Final, validation validation will to determine determine that the logic logic of that the reasonableness of the results can be verified, verified. comport to the guidelines cited and Entergy Entergy procedures referenced in PSA the cutset or sequence is Palisades indicated that this review is is required but not Notebook Notebook NB-PSA-CC NB-PSA-CC [6].

correct.

correct. complete. This This finding is being written written against allall of the QU-D updated flooding PRA The updated PRA is is fully documented documented in in QU-D supporting supporting requirements requirements as as well as some QU-F requirements. EA-PSA-INTFLOOD-13-06 Rev.

EA-PSA-INTFLOOD-13-06 Rev. 00 [43].

[43].

requirements. Palisades needs needs to complete the formal formal review ofof accident accident sequence sequence quantification quantification results and make results and make modifications as as needed needed to to address issues Finding Open Finding Open issues found found in in that that review.

review. The The final final results results should should then bebe documented documented in in the the corresponding corresponding notebooks.

notebooks.

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50.55a(a)(3)(i)

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

F&O#

Finding or Finding or ASME Reg.

ASME Reg. Guide Guide 1.200 1.200 (Supporting (Supporting Finding Description Finding Description (summary (summary discussion) discussion) Disposition Disposition Suggestion Suggestion Category IIII Text Category Text Requirement)

Requirement)

SC-A5-01 SC-AS-01 Suggestion Suggestion SPECIFY an SPECIFY an appropriate appropriate Palisades uses Palisades uses 24 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> hours as the default mission as the time mission time Palisades has Palisades has re-evaluated re-evaluated the the data data analysis analysis and and the the time time dependent dependent mission time for the mission the all sequences that end for all end in stable end in aa stable end state. This state. This models for the models the treatment treatment ofof LOOP LOOP events.

events. The The modeling modeling aspects include include modeled accident modeled accident can be can be potentially potentially oveny overly conservative conservative for for some some the time LOOP recovery, time of LOOP recovery, the time of of onsite onsite power power system recovery, recovery, sequences. For sequences. For sequences sequences sequences suchsuch as LOOP LOOP sequences sequences when when power power is is mission time, EDG mission EDG time, and the coping timetime between between the time of an an SBO SBO event event in which in which stable stable plant plant hours. A recovery not recovered by 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. recovery factor and the time when electric power must must be recovered to prevent core conditions have conditions have been considering the convolution of of EDG EDG FTR FIR with offsite damage. In In addition addition to analyzing analyzing these these interactions, interactions, the the August 2003 2003 achieved, USE a minimum minimum power was used power used but but did not not account account for increased blackout event is evaluated northeast blackout evaluated inin the data analysis analysis as well.

mission time of 24 hr.

mission time hr. time for recovery as as aa function function ofof the time that the times for individual Mission times could run before failure EDG could The approach used used to develop appropriate recovery recovery factors inin the analysis Mission SSCs that function during as follows:

is as SSCS the accident accident sequence 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 the -* A best estimate model is adopted based on the NUREG/CR-6890 curve than 24 be less than 24 hr, hr, as long overestimates the importance of some events.

be 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 and operator operator actions -* To account for uncertainty in the model and curve fitting and to account SSCs and credit the increased time available for recovery of are modeled modeled to to support the the for the sparcity of data, an upper bound model and lower bound model are offsite power as a function of how long the EDG runs were also devised. These upper and lower bound models are full sequence mission time. before failure.

example, if following a For example, established by the following relationships:

LOCA, low low pressure *The upper bound, best estimate, and lower bound models are

-The available for 11 injection is available assumed to coincide at t=O t=0 hour, after hour, after which recirculation is required, the recirculation

• At t=1

-At t=100 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 mission time for LPSI may be 11 hour1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> and the mission lower than the best estimate model in terms of the cumulative non- non time for recirculation recirculation may be exceedance probability of the time to recover offsite power.

time 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />. For 23 For sequences in *At

-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 notnot be achieved by 24 lower than the best estimate model in terms of the cumulative non- non hr using the modeled plant exceedance probability of the time to recover offsite power.

equipment equipment and and human actions, PERFORM PERFORM *The

-The factor difference between the upper upper and and lower bounds and the additional additional evaluation or or best estimate is is permitted to grow grow in in a log-linear log-linear fashion from 0 to modeling by modeling by using using anan 10 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, hours, from 10 10 to 20 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, hours, and and beyond 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

hours. This This means appropriate appropriate technique. that thethe curve fit uncertainty uncertainty factor is allowed to to grow uniformly 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 /> hour period.

techniques include:

include: (a)

(a) *- In In the the uncertainty uncertainty analysis, aa discrete discrete distribution distribution over over the three assigning an assigning an appropriate curves is is used used with 10%

10010 probability probability assigned to the Upper Upper and and plant damage plant damage state for the the Lower Bound Bound Models, and and 80% probability probability to to the the Best Best Estimate Estimate sequence; (b) (b) extending the the Model.

Model.

mission mission time, and and adjusting the affected the affected analyses, analyses, to to the the Suggestion Suggestion Resolved Resolved point at which point at which conditions conditions can can bebe shown shown to to reach reach ----

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

F&O# Finding or Finding or ASME Reg.

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

Disposition Disposition Suggestion Suggestion Category IIII Text Category Text Requirement)

Requirement) acceptable values; or acceptable or (c)

(c) modeling additional modeling additional system system recovery or recovery or operator operator actions for the sequence, in for in accordance with accordance with requirements stated requirements stated in in Systems Analysis (2-2.4)

Systems (2-2.4) and Human Reliability and Human Reliability (2- (2-2.5) to demonstrate that that aa successful outcome successful outcome is achieved.

achieved.

SC-B5-01 SC-BS-01 Suggestion Suggestion CHECK the 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 and acceptability of the and consistent, there was no documented evidence [1 1]. This section describes a comparison of the Palisades success

[11].

results of results the of the that they had been checked against generic genenc or other criteria to some comparable event tree headings developed for Waterford thermal/hydraulic, structural, thermallhydraulic, plants. Palisades did provide some documentation on 3, which is a similar Combustion Engineering designed PWA. 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 support the success criteria.

success criteria. Group guidance Group guidance but but there there was no single, centralized was no Examples of of methods methods to of documentation Suggestion Resolved Examples to set of set documentation to to demonstrate demonstrate how Palisades achieve this achieve this include:

include: (a) (a) met the met comparison requirement the comparison requirement ofof the SA.SR. Palisades comparison with comparison with results results of of needs to needs provide documentation to provide documentation of of the the comparison to the same the analyses same analyses other generic or similar plants or provide a set of performed for performed for similar similar plants, plants, references to references to other other documents documents that that support this accounting for accounting for differences differences in in requirement.

requirement.

unique plant unique plant features features (b) (b) comparison comparison with with results results of of similar analyses similar performed analyses performed with other plant-specific with other plant-specifiC codes (c) codes (c) check check by other by other means means appropriate appropriate to to the the particular particular analysis analysis SC-C2-01 SC-C2-01 Suggestion Suggestion DOCUMENT DOCUMENT the the processes processes LOCA break sizes LOCA break sizes are given in detail. However, However, the The primary technical basis reference is included in Attachment 2 of NB-used used toto develop develop overall overall PRA PRA traceability ofof the the references provided provided for where and PSA-lE Rev. 4. The thermal hydraulic basis was developed PSA-IE Rev. developed in in PLP0247-success criteria success criteria and and the the how how these break sizes were determined sizes were determined is is difficult difficult to 07-0004.01 R2, "Palisades Nuclear R2, Palisades Nuclear Plant Plant Thermal Hydraulic Hydraulic MAAP supporting engineering supporting engineering follow to the follow to the ultimate ultimate basis.

basis. Based on discussion with with Calculations Calculations" [14].

[14]. Additional description description and technical basis basis is is contained bases, including bases, including thethe inputs, inputs, the lead lead PRA Engineer, Engineer, a reference is is available for in in EA-PSA-LOCA-IE-12-02, Initiating "Initiating Event Event Frequencies for Loss of methods, and methods, and results.

results. For For these break sizes.

these break sizes. Coolant Coolant Accidents for the the Palisades Nuclear Nuclear Plant Plant Probabilistic Probabilistic Risk example, this example, this documentation documentation Documentation Assessment Assessment" [30].

[30]. References to to these documents documents were were added added to to Section Section Documentation only. only.

typically includes:

typically includes: (a) (a) the the 5.0 of NB-PSA-ETSC 5.0 of NB-PSA-ETSC [1 1].

[11].

definition definition ofof core core damage damage Include Include aa reference in in the the success success criteria criteria notebook notebook Page 41 Page 41 of of 55 55

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

F&O#

(Supporting Finding or ASME Reg. Guide 1.200 Finding Description (summary discussion) Disposition Suggestion Category IIII Text Requirement) 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 available mitigating systems andand human actions for each accident initiating group modeled in the PRA (g) the basis for establishing the time available for human actions actions (h) descriptions descriptions of processes usedused to define define success criteria criteria for grouped initiating events or accident sequences sequences SC-C3-01 Finding Finding DOCUMENT the sources of Some Some Calculations associated with success criteria criteria With regard to success success criteria, the technical reference is documented documented in model uncertainty and and are not not in the the Palisades formal document document control control the event tree and and success criteria criteria notebook:

notebook: PLP0247-07-0004.01 RO, related assumptions (as (as system.

system. In basis for the LOCA size In addition, the basis Palisades Nuclear Plant Plant Thermal Thermal Hydraulic Hydraulic MAAP MAAP calculations calculations (R-1551).

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

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

F&O# Finding or Finding ASME Reg. Guide 1.200 (Supporting I

Finding Description (summary discussion) Disposition Disposition Requirement) Suggestion Category II Text Requirement) identified in aU-E1 QU-El and au-QU- ranges are not included. The issue regarding the ranges basis regarding LOCA size and frequency Additional discussion and basis 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 F&0 SC-C2-01 in suggestion F&O SC-C2-01.. 0,Calculation 0, Initiating Event Frequencies in Accordance with CEOG "Calculation of Initiating criteria. Standards..

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 EA-PSA-IE-OO-OO10 License Amendment Request for NFPA 805 or a NB-PSA-ETSC to improve the discussion regarding the basis basis for the Therefore, some calculations are not power uprate). Therefore, 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 the PRA results. Technical bases for the size ranges NUREG/CR-5750 for use particularly in work associated with from NUREGlCR-5750 10CFR5O.46 but with applicability for other risk-informed applications such 10CFR50.46 are not included in the success criteria definitions.

concem that the LOCA frequencies in as the PTS project. There was a concern 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.

5.77E-04/yr. In summary the Palisades LOCA frequencies are well documented 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 Finding SY-A13-01 SY-A13-01 Finding INCLUDE those failures that Currently a flow diversion diversion pathway is modeled for the Gates FLW-DIV-P54B&C-INJ, FLOW DIVERSION TRHOUGH "FLOW DIVERSION TRHOUGH P-54B can cause flow diversion diversion Containment Spray pumps pumps failing due to a diversion diversion AND P-54C DURING INJECTIONINJECTION MODE, MODE", FLW-DIV-P54A&B-INJ FLOW "FLOW pathways that result in in through a failed other Containment Spray Spray pump with a DIVERSION TRHOUGH TRHOUGH P-54A AND P-54B DURING DURING INJECTION INJECTION failure to meet meet the system failed outboard check valve. Although this is a valid MODE, MODE", and FLW-DIV-P54A&C-INJ FLOW "FLOW DIVERSION DIVERSION TRHOUGH TRHOUGH Page 43 of 55

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

F&O# Finding or ASME Reg.

Reg. Guide 1.200 (Supporting (summary discussion)

Finding Description (summary Disposition Suggestion Category II Text Requirement) criteria, success criteria. flow diversion pathway during the injection mode of P-54A AND P-54C DURING INJECTION MODE" MODE were added to the PRA operation, it is not a flow diversion pathway during model (See Attachment E Section 7.3.2).

diverted flow would be recirculation since the "diverted" 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 mode of operation. When true, the flow diversion results in no flow from Spray (CS) flow is supposed to go anyway. Because I

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 tolthrough 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 similar divisions (such as train A of AFW and train A To "To evaluate coincident unavailability, all the unavailability data was of redundant equipment compiled, and coincident events were marked. Coincident unavailability when this is a result of of HPI) but does not include events that might occur planned associated with coincident unavailability was considered for each train (i.e., (i.e., 2 or more more train A components components OOS at planned activity (see DA- unavailability of of multiple C14). trains of different the same time), and for both both trains (i.e., 1 or more Train A components (i.e., 1 different systems that belong belong to opposite division (such as train A of AFW and train B of HPI). OOS at the same time as as 11 or more more Train B B components). In In addition to reviewing the maintenance rule unavailability data for coincident unavailability, the risk management work week reviews from the LAN were unavailability, also downloaded and reviewed.

Potential unavailability between systems involving Potential involving opposite opposite divisions due to planned activities activities isis not The following follOwing identifies the equipment associated with each train:

included in the model and and may result in in non non-

• • Train A equipment: C-2A & C-2C, C-6B, C-SB, ED-15 ED-15 & ED-17, K-6A, P-52C, K-6A, P-52C, conservative results.

P-54B & P-54C, P-55C,

& P-54C, P-55C, P-56A, P-5SA, P-66B, P-66B, P-67B, P-S7B, P-7B, P-7B, P-8A & P-8B, and PRy-i PRV-1042.042.

Include events in the model that address coincident Include coincident ** Train BB equipment:

equipment: C-2B, C-6A, C-SA, ED-16 & ED-18, K-6B, P-52B, P-52B, P-54A, unavailabilities associated associated with train A of oneone system system P-55A P-55A & P-55B, P-56B,

& P-55B, P-56B, P-66A, P-67A, P-7A P-7A & 7C, 7C, P-SC P-8C and PRy-i PRV-1043.

043.

with train BB of of another, redundant systems systems due due to Plant experience experience showed showed that in in most most cases onlyonly one piece of equipment equipment Page 44 44 of of 55

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

F&O# Finding or Finding or ASME Reg.

ASME Reg. Guide Guide 1.200 1.200 (Supporting (Supporting Finding Description (summary Finding Description (summary discussion) discussion) Disposition Disposition Suggestion Suggestion Category IIII Text Category Text Requirement)

Requirement) planned activities planned activities (if (if the experience experience shows shows any exist).

exist). from aa train from train is is removed removed from service service at time. A at aa time. review of A review of the the three plus plus unavailability data years of unavailability data showed thatthat there there was was limited, limited, repetitive repetitive coincident unavailability; coincident unavailability; most most cases involved involved onlyonly two two components, components, and occurred only occurred once in only once in the three year year data window.

There were, however, a few cases in which plant experience showed that two components were recurrently removed from service at at the same time.

coincident unavailability was modeled; the following In these cases, coincident identifies the combinations of equipment for coincident unavailability: 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-67P-67A;A; and

7. P-56A and P-56B.

Basic events were developed for items 1-7 1-7 above and documented in Attachment 12, Table 12.1 of Reference [5].

Coincident unavailability included only the time that both components were simultaneously unavailable. If one component was unavailable for hour, the hour was used an extra hour, used inin the individual unavailability. Once coincident unavailability's unavailabilitys were were calculated, the times were subtracted from the individual unavailabilitys unavailability's to avoid double counting."

counting.

Finding Finding Resolved SY-B3-01 SY-B3-01 Finding Finding ESTABLISH common cause ESTABLISH Common cause failures as a whole are are modeled A full evaluation of of this finding is is presented in Attachment 11 of of Reference failure groups groups byby using a correctly and consistently. However, However, the modeling modeling ofof [9].

logical, logical, systematic systematic process the HPI, HPI, LPI, and common line line check valves is that Examination Examination of cut sets sets that include CCF CCF of of in-series components reveals that considers similarity similarity in in producing non-minimal and and potentially non-valid (a) service conditions (b) cutsets. that there there are no no non-minimal non-minimal cut sets.

sets.

(b) environment environment (c) design or (c) design or Treating Treating in-series in-series HPSI HPSI andand LPSI LPSI valves as as independent independent (incorporating (incorporating Because Because of of the safety significance the safety significance ofof the the LPI LPI and and HPI HPI manufacturer manufacturer (d) (d) the the CCF CCF portion of of the the valve valve failure failure in in the the failure failure probability probability forfor each each systems, the non-minimal systems, the non-minimal and and non-valid cutsets cutsets are are maintenance maintenance JUSTIFY JUSTIFY the valve),

valve), asas appears appears to be suggested to be suggested by by this finding, finding, turns turns out out toto be be the basis overestimating overestimating thethe risk associated associated with those failures.

with those failures.

basis for selecting selecting common more more conservative conservative approach.

approach.

cause component component groups.

groups. Review the common Review the common causecause modeling modeling of of components components Candidates The The Palisades Palisades approach approach produces produces realistic realistic andand valid valid results.

results.

Candidates for for common common in the PRA in the PRA model, model, especially especially ofof the the valves valves in in series series cause failures cause failures include, include, for and revise and revise the the model model as as appropriate.

appropriate. Alternatively, Alternatively, The modeling of The modeling of common common causecause failures, failures, as as applied applied in in the the Palisades Palisades PRA, PRA, Page Page 45 45 of of 5555

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

F&O# F d Finding or ASME Reg.R G *d e 1.200 Guide 1 200 (Supporting categorre Finding Description (summary discussion) Disposition Suggestkn Suggestion Category II Text Requirement)

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 The approximation suggested by this finding is, in fact, a more valves (f) check valves (g) (ri) 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) (j) circuit The approximation used in the Palisades PRA, namely, using the "total" total breakers independent failure rate without correcting by failure rate to represent the "independenr the factor of (1-beta), also does not introduce significant conservatism in the results.

concerns expressed by this finding do not appear to be Therefore, the concems correct, and modeling or quantification changes are not considered necessary.

Finding Resolved SY-B4-01 SY-84-01 Suggestion INCORPORATE common Because the CCF modeling approach for CCGs An evaluation of this peer review team suggestion was perfonned performed in cause failures into the greater than 5 is bounding, it is recommended that the Attachment 4 of EA-PSA-RG1.200F&O-10-01 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

-The model used for data anal y sis tailing due to common cause has a value that is higher (by a factor of failing analysis.

. more than 50) than the factor that is calculated using the explicit multiple Greek letter (MGL) approach.

• Atthe

-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 global components, and vice versa. This is the expected result.

• At the system level, the quantification of a system

-At "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

-For overall results of the PRA, it likely would need to have an effect on multiple redundant systems. Such global CCF events may exist exist in in the Palisades Palisades PRA (e.g., station power transformers, transfonners, sequencers) and should be be examined for potential potential further refinement of of the CCF when they impact impact the results results of of an application significantly.

Page 46 of 55

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

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

F&O# F d Finding or ASME Reg.

ASME A G *d e 1.200 Guide 1 200 (Supporting (Supporting Finding Description (summary Finding Description (summary discussion) discussion) Disposition Disposition SuggesUon Suggestion Category IIH Text Category Text Requirement)

Requirement)

The impact The impact of the the bounding bounding approach approach is is to produce produce results results that are are conservative, with conservative, with the the amount of of conservatism conservatism being being aa function function of the component group component group and and its its MGL factors.

factors.

Suggestion Resolved Suggestion SY-B5-01 SY-B5-01 Finding Finding ACCOUNT explicitly for the There is an apparent errorerror in the EDGEDG failure to run not considered a finding. SWS start failures are captured This is not captured under under modeled system's modeled systems logic: this logic does not account for the SWS pump the diesel failure to start gates. Start, load/run load/run and runrun failures are all dependency on support dependency failures to start. When the PRA Group was shown the captured under 'OR'OR gates so the logic logic is equivalent.

systems or interfacing apparent error, they admitted that it was an error and systems in the modeling The PRA model Release 2b cutsets properly account for diesel run and systems that they had also identified it in their Self-process. This may be service water pump start failures.

process. Assessment. The model was corrected while the accomplished in one of the Review Team was on-site, but a review of the This issue was noted under supporting requirement aU-D5 QU-D5 in the Reg.

following ways:

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 fault 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 dependencies to start, the diesel generator fail to run should be 1.0.

appropriate event event or gate in Finding Resolved appropriate the support system the system fault tree; (b) for the linked event SWS pump failures to start are valid contributors to tree approach, by using tree EDG failure. The model should account for these event tree event tree logic logic rules, or contributors and the diesel generator failures need to calculating a a probability for for be adjusted to account for the availability of SWS each split fraction conditional on the scenario conditional definition.

definition. These specific failures should be incorporated into the fault tree model. And, given the similarity of of this finding with Finding SY SY-B5-02,

-B5-02, itit is recommended that a systematic review of otherother potentially risk important dependencies be performed.

SY-B5-02 SY-B5-02 Finding Finding ACCOUNT ACCOUNT explicitly for the Potentially risk-significant manual valves were Potentially were Per supporting supporting requirement SY-A15SY-A 15 [1]: A component may be excluded modeled modeled systems system's excluded from the model withoutwithout explanation.

explanation. Their Their from the system model if the total failure probability of the component dependency on on support exclusion should be be based on on SR SY-A15 SY-A 15 screening failure modes modes resulting in in the same effect effect on system system operation operation is at least least systems or or interfacing interfacing criteria. For For example, example, manual manual valves valves in in the two orders ofof magnitude lowerlower than than the highest failure probability of the the systems systems in in the the modeling modeling Containment Containment SpraySpray system system flow paths were not paths were not other other components components in in the the same same system system train train that that results results in in the the same same effect effect process.

process. This This may may be be modeled.

modeled. on on system system operation.

accomplished accomplished in in one one of of the following ways:

ways: (a)(a) for the the fault tree linking fault tree linking approach approach ItIt was was noted that some noted that some of of these these manual manual valves valves areare The The valves valves described described inin finding finding SY-B5-02 SY-B5-02 in in the the containment spray spray system system by by modeling modeling the the actually actually depicted depicted on the simplified on the system drawings, simplified system drawings, are are normally normally locked open manual locked open manual valves.

valves. The The Palisades Palisades PRA PRA has has dependencies dependencies as as aa link link to to an an but they are but they not labeled.

are not labeled. To To avoid avoid confusion, confusion, itit is is assumed assumed that that random random failure or plugging failure or plugging ofof locked locked open open manual manual valves valves Page Page 47 47 of of 55 55

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Table A2.5-1, March 2010 Full Power Internal Events Peer Review Report Findings and Resolutions r:u~:Orting orting ~inding ~r ~S~E Re~*TG~de 1.200 1.200 Rnding Finding Description 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.

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.

included, success criteria notebook [11], [1 1], and appropriate system notebooks.

rules or event tree logic rules, This provides a quick indication of what components ' .

calculating a probability for are physically present but not explicitly modeled. The assumptions states:

The assumptions states.

each split fraction NUREG CR-6928 (January 2007) Table 5-1, provides data for manual "NUREG conditional on the scenario valve failure to open and failure to close. Failure to remain open is not definition. significant.

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! 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 SY 5 or include them in the model.

-A 15 keys issued from the control room and systems with locked open valves I normally in operation or are frequently tested to meet technical are either nonnally I specification requirements. A mispositioned or plugged valve would likely be detected as part of plant operator rounds, testing.

testing, check lists, 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 plugging, testing, of these valves, 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."

maintenance.

With respect to manual valve misalignment.

misalignment, a scoping analysis was performed as documented perfonned documented in the Palisades HRA notebook NB-PSA-HR Volume 2. 2, "Palisades Palisades Pre-initiator Human Human Error Evaluation" Evaluation [8]. The results of this scoping demonstrated that a number of manual valves are susceptible susceptible to mispositioning with a non-significant failure probability. The basic events developed and and scoping methodology are presented presented in that document.

document.

A sampling of locked open and non-locking manual valves were evaluated A

from the auxiliary feedwater, feedwater. shutdown cooling, cooling. and atmospheric dump dump valve fault trees to provide provide validation that assumption A-0047 is applied consistently to all system fault trees. No discrepancies were found.

Finding Finding Resolved SY-B1 1 -01 SY-B11-01 Finding Rnding MODEL the ability MODEL ability of the The current model model for the supplemental diesel diesel model, model. A A new new fault tree was created, created. PNOSGPWR PNOSGPWR NO "NO SAFEGUARDS SAFEGUARDS POWER POWER available inventories of air, however, however, is not correct as CB 152-203 not completely correct 152-203 TO TO SAFEGUARDS SAFEGUARDS BUS BUS" that models failure of buses 11C, C. 1 D. or 1 D, or 11EE and power, power. and and cooling cooling to should be fails

-fails to remain closed closed- instead of of Fails "Fails to failure of of their respective breakers that tie tie them to the safeguards safeguards bus to to support the mission time. remain open, and failure of the A14 safeguards open-. and safeguards bus bus open.

open.

needs_to_be_added_to_the_model_as_a_reason_the needs to be added to the model as a reason the ___ _ ~_ _ _ _ __

Page 48 of 55

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

Table A2.5-1, March 2010 Full Power Internal Internal Events Peer Review Report Findings and Resolutions F&O# Finding or ASME Reg. Guide 1.200 (Supporting Finding Description (summary discussion) Disposition Suggestion Category II Text Requirement)

Supplemental DG fails to provide power to the 11 D Supplemental This fault tree was placed under gates:

Safeguards bus.

• • PNOSGPWR1 D, NO SAFEGUARDS POWER TO BUS 11 D

• PNOSGPWR1C, NO POWER FROM SAFEGUARDS BUS TO BUS The 152-203 CB is modeled under another portion of 1C power to the 11D IogkD for power the logic Saf eguards bus. USing D Safeguards Using PNOSGPWR1 E, NO SAFEGUARDS POWER TO BUS 1E

• • PNOSGPWR1E, 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 adequately captured in the model.

not being adequat~ly 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 11 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 11A A bus itself to the appropriate bus.

model.

DG-NSR-START1D-03, Under gate DG-NSR-START1 CIRCUIT BREAKER FAILURES" D-03, "CIRCUIT 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 152-203 152-203 FAILS TO CLOSE DG-NSR-RUN1D-03, Under gate DG-NSR-RUN1 CIRCUIT BREAKER FAILURES" D-03, "CIRCUIT FAILURES added basic events:

• ACP-C2MC-152-203, CIRCUIT BREAKER 152-203 152-203 FAILS TO REMAIN CLOSED Under gate DG-NSR-START1C-03, "CIRCUIT CIRCUIT BREAKER FAILURES" FAILURES added basic event:

• ACP-C2MB-152-403, ACP-C2MB-152-403, NSR EDG OUTPUT BREAKER 152-403 152-403 FAILS TO CLOSE These logic changes capture all of the appropriate appropriate breaker failure modes related to the non-safety related emergency diesel generator and the safeguards bus.

Finding Resolved SY-B12-01 SY-B12-01 Finding Finding DO NOT USEUSE Palisades did did not not model HVAC HVAC for the control room or or The basisbasis for excluding control control room HVAC HVAC from the full power internal intemal proceduralized proceduralized recovery the cable spreading room based based on operator actions on operator actions events events model model was strengthened include other aspects strengthened to include aspects in addition to actions actions as the the sole basis basis for to implement to implement alternate altemate cooling strategies such as operator operator actions. This evaluation was was fully documented documented inin NB-PSA-ETSC eliminating a support system opening doors or using using a proposed portable portable exhaust [1 1]. The conclusion summary

[11]. summary states:

from the model; however, however, fans. (See pages 17 17 and 24 of attachment attachment 8 to NB-Page 49 of 55

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

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

F&O# Finding or Finding ASME Reg.Beg. Guide 1.200 1.200 (Supporting Finding Description (summary discussion) Disposition Suggestion Category II Text Requirement)

INCLUDE these recovery PSA-ETSC r01).

rOl). However, the operator actions to Control room cooling in the Palisades internal events PRA is not actions in the model alternate actions were not included in implement the altemate considered an issue based on the following:

quantification. For example, the models. There was never an intent to model the it is not acceptable to not operator actions given that past analyses has shown

• the high design temperature limits of the major control room components, model a system such as that both rooms can survive a loss of HVAC. It is 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 HVAC AC systems for the cable room during such an event, spreading room and control room, or model the alternate cooling operator actions to implement altemate TMMs are not credited in the EOPs,

• the TMM's 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 internal events model.

an initiator or as a support system for the intemal 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 rooms heat load based on Systematic Evaluation Program (SEP) Topic IX-5 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 CALC-455-001-DC2 [13] was then performed to evaluate evaluate all all cable spreading room equipment modeledmodeled in the PRA under these conditions.

conditions.

The analysis analysiS conservatively assumed that the room was at the peak calculated temperature temperature of 122°F DC) 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 (50°C) 122DF (50 the transient. An evaluation of eqUipment equipment qualification qualification reports, and vendor data, was then performed which concluded that reasonable assurance of operability operability is assured for all equipment at an elevated elevated ambient of 122°F 122DF for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Based Based onon the conclusions conclusions of these analyses, analyses, ventilation ventilation to the cable spreading area is not not explicitly modeled modeled asas failure to re-establish ventilation does does not result in not result equipment failure prior in equipment prior to to the PRA 2424 hour0.0281 days <br />0.673 hours <br />0.00401 weeks <br />9.22332e-4 months <br /> hour mission time.

Attachment 88 of NB-PSA-ETSC 111] [11] has has been been updated

~dated to reflect these these -

Page 50 of 55

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

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

F&O# Finding or Finding ASME Reg. Guide 1.2001.200 (Supporting Finding Description (summary discussion) Disposition Suggestion Category II Text Requirement) conclusions.

Finding Resolved SY-B1 2-02 SY-B12-02 Finding DO NOT USE Detailed analysis of systems/component Dependency The basis for excluding control room HVAC from the full power intemal internal proceduralized recovery on HVHVAC/ventilation 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:

[11).

from the model; however, the basis for not modeling the HVAC within the INCLUDE these recovery system notebooks for the control room and cable internal events PRA is not Control room cooling in the Palisades intemal actions in the model spreading room. following:

considered an issue based on the following:

quantification. For example,

• the high design temperature limits of the major control room SWS pump failures to start are valid contributors to it is not acceptable to not components model a system such as EDG failure. A review of the affected cutsets still has a cutset with a diesel generator run failure in the

• the general conservative modeling assumptions employed throughout HVAC or CCW on the basis same cutset as the SWS pump failure to start. Given the EA-APR-95-023 Rev. 1 1 analysis that there are procedures for failure of the SWS pump to start, the diesel generator dealing with losses of these

• the philosophy of the operators with respect to remaining in the control systems. fail to run should be 1.0. The model needs to account systems. room during such an event, for these and similar dependencies.

dependencies.

These specific failures should be incorporated into the

• the TMM's TMMs 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 SY-B5-02, finding with Finding SY 02, it is recommended that studies a systematic review of other potentially risk important dependencies be performed. Therefore it is considered unnecessary to model either loss of HVAC as an initiator or as a support system for the intemal 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]. [12). The analysis approach was to establish the room's rooms heat load based on Systematic Evaluation Program lX-5 (Phase II cable spreading room loss of HVAC testing)

(SEP) Topic IX-5 data by modeling the test boundary conditions in detail and iterating on room heat generation until the test results were mimicked mimicked byby 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]

CALC-455-001-DC2 [13) was then performed to evaluate all cable spreading room equipment modeled modeled in in the PRA under under these conditions.

conditions.

The analysis analysis conservatively assumed assumed that the the room was at at the peak calculated temperature of 122°F 122°F (50°C)

(50°C) for the the entire 48 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> hour duration duration of the transient. An evaluation of the of equipment qualification qualification reports, and vendor vendor data, data, was then then performed which concluded that reasonable assurance of Page 51 of 55

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

Table A2.5-1, March Table A2.S-1, March 20102010 Full Full Power Power Internal Internal Events Events Peer Peer ReviewReview Report Report Findings Findings and and Resolutions Resolutions F&O #

F&O#

Finding or Finding or ASME Reg.

ASME Reg. Guide Guide 1.200 1.200 (Supporting (Supporting Finding Description Finding Description (summary (summary discussion) discussion) Disposition Disposition Suggestion Suggestion Category IIII Text Category Text Requirement)

Requirement) operability is operability is assured assured for all equipment for all equipment at at an an elevated elevated ambient ambient ofof 122°F 122°F hours.

for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Based the conclusions Based on the conclusions of of these these analyses, analyses, ventilation ventilation toto the cable cable spreading area is not not explicitly explicitly modeled modeled as failure toto re-establish ventilation does not not result result in in equipment failure failure prior prior to the PRA 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> hour time.

mission time.

NB-PSA-ETSC [11] has Attachment 8 of NB-PSA-ETSC has been been updated to reflect reflect these these conclusions.

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' 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- EA-PSA RG1 .200F&O-1 0-01 [9].

RG1.200F&O-10-01 Finding Resolved SY-B14-01 SY-B14-01 Suggestion Suggestion IDENTIFY SSCs 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 conditions beyond their Sump Blockage potential. A discussion of the sump NB-PSA-DA [5]). [5]). Description of sump strainer and discussion of sump environmental environmental qualifications, 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 (NB-PSA-SY-SSS -SSS [21], Sections Sections 1.0, 1.1 and 2.12).

1.0, 1.1 failures failures of multiple SSCs was not found in the associated fault tree model.

that result from operation operation in Suggestion Resolved Note: independent sump blockage events are adverse conditions.

these adverse conditions. included inin the model.

Examples Examples of degraded degraded environments environments include include (a)

(a)

LOCA LOCA inside inside containment Because Because of the the significance significance and impact impact of the sumpsump with with failure failure of containment blockage potential, impact of this issue should be potential, the impact heat removal (b) safety discussed in in the system notebook and and included in in the relief valve valve operability operability (small (small model as as appropriate. (Note:(Note: Palisades Palisades did did identify identify LOCA, drywell LOCA, drywell spray, spray, severe this this issue in their self-assessment but issue in but itit remained accident) accident) (for (for BWRs)

BWRs) (c) (c) unresolved unresolved at at the the time time of the peer of the peer review.).

reView.).

steam line breaks steam line breaks outside outside containment containment (d) debris that (d) debris that could could plug plug screens/filters screens/filters (both internal (both and external internal and external to to -----

Page Page 52 of of 55 55

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

Table A2.5-1, Table A2.5-1, March March 20102010 Full Full Power Power Internal Events Peer Internal Events Peer Review Review Report Report Findings Findings and and Resolutions Resolutions F&O #

F&O# Finding or Finding or A S ME Reg.

ASME 1 200 G Ui*d e 1.200 Reg. Guide (Supporting (Supporting Finding Description Finding Description (summary (summary discussion) discussion) Disposition Disposition Suggestion Suggestion Category IIII Text Category Text Requirement)

Requirement) the plant) the plant) (e) heating of (e) heating of the water supply water supply (e.g., BWR (e.g., BWR suppression pool, PWR suppression PWR containment sump) that could affect could affect pump pump operability NPSH for pumps (f) loss of NPSH (g) steam binding of pumps.

pumps.

INCLUDE operator interface INCLUDE dependencies across dependencies systems or trains, where applicable.

SY-B15-01 SY-B1S-01 Suggestion Suggestion INCLUDE operator 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 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.

CCS-Door-167B B. (EA-PSA-CCW-HELB-02-1 7 [22]). Updated Sections 2.6 and 2.7 of the (EA-PSA-CCW-HELB-02-17 applicable.

applicable. Attachment B in tum pointed to a file Entitled CCC turn pOinted cooling system notebook, NB-PSA-SY component COOling NB-PSA-SY-CCS

-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.

only. Palisades needs to clean up these references.

Page 53 Page 53 of of 55 55

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

A2.6 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 Industry Peer Review the Probabilistic Safety Analysis (PSA) against the Combustion Engineering Owners Group checklists, RIE 2000-02, CE-NPSD-1194-P (CEOG) PSA checklists', CE-N PSD-1 194-P Task 1037.1037.

[3] PALISADES GAP ANALYSIS REVIEW AND ERIN Engineering and Research Inc., "PALISADES UPDATE, P0495060007-2711-061215, October 2004.

UPDATE",

[4] From David Finnicum to Bradford Grimmel, "RG RG 1.200 1.200 PRA Peer Review Against the ASME PRA Standard Requirements For The Palisades Nuclear Power Plant Probabilistic Assessment, LTR-RAM-II-10-015, March 12, 2010.

Risk Assessment",

[5] Palisades PSA Notebook NB-PSA-DA Rev. 6, "PalisadesPalisades PSA Data Notebook".

Notebook.

[6] Palisades PSA Notebook NB-PSA-QU Rev. 3, "Quantification Guideline.

Quantification Guideline".

[7] EA-PSA-1 999-010 Rev 0, "Palisades EA-PSA-1999-010 Palisades PSA Bayesian Update".

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 EA-PSA-RG1.200F&O-10-01 Resolution of Reg. Guide 1.200 October 2009 Full Observations.

Power Internal Events Peer Review Findings and Observations".

[10] Initiating Event Notebook".

Palisades PSA Notebook NB-PSA-IE Rev. 4, "Initiating Notebook.

[11] Palisades PSA Notebook NB-PSA-ETSC Rev. 3, "Event Event Trees and Success Criteria".

Criteria.

[12] EA-PSA-GOTHIC-CSRHEATUP-09-09 Rev. 0, "GOTHIC GOTHIC Cable Spreading Room Heat-Up.

Heat-Up".

[13] Evaluation of Equipment in the CSR when Exposed to CALC-455-001-DC2 Rev. 0, "Evaluation Hours.

Elevated Temperatures for 48 Hours".

[14] Palisades Nuclear Plant Thermal Hydraulic MAAP PLP0247-07-0004.01 Rev. 2, "Palisades Calculations..

Calculations"

[15] CALC-455-001 -DC1 Rev. 0, "Survivability CALC-455-001-DC1 Survivability of Equipment inside Containment Following a LOCA/MSLB.

PRA LOCAlMSLB".

[16] Palisades Internal Events and Fire PRA Model".

EA-PSA-FPIE-FIRE-12-04 Rev. 0, "Palisades Model.

[17] Palisades PSA Notebook NB-PSA-CC Rev. 1, "PSA PSA Model Configuration Control".

Control.

[18] Palisades PSA Notebook NB-PSA-SA Rev. 1, "RG RG 1.200 PRA Self-Assessment Against Requirements.

the ASME PRA Standard Requirements".

[19] EA-PSA-IE-00-001 0 Rev. 0, "Calculation EA-PSA-IE-00-0010 Calculation of Initiating Event Frequencies in Accordance Standards.

with CEOG Standards".

[20] Not Used

[21] Palisades PSA System Notebook NB-PSA-SY-SSS Rev. 1, "SIRW SIRW Tank and System.

Containment Sump Suction System".

[22] EA-PSA-CCW-HELB-02-1 7 Rev. 0, "Evaluation EA-PSA-CCW-HELB-02-17 Evaluation of the Impact of a High Energy Line Break in CCW Room with either Door 167 to 590 Corridor Auxiliary Building or 167B 1 67B 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". Open.

[23] Palisades PSA Notebook NB-PSA-SS Rev. 2, "Palisades Palisades Safe and Stable States".

States.

[24] Nuclear Management Company, "Update Update of Palisades CDF Model - PSAR2b to EA-PSA-PSAR2c-06-1 0, Rev. 0, June 2006.

PSAR2c, Calculation No. EA-PSA-PSAR2c-06-10, PSAR2c",

[25] Palisades PSA Notebook NB-PSA-HR Rev. 4, Palisades Human Reliability Analysis Notebook Volume 1 1 (Post Initiator Operator Actions).

[26] Level 2 Notebook".

Palisades PSA Notebook, "Level Notebook.

[27] Palisades Pressurized Thermal Shock (PTS) Probabilistic Risk Assessment (PRA),',

"Palisades (PRA),

ADAMS Accession number ML042880473, October 6,2004. 6, 2004.

[28] Palisades PSA Notebook NB-PSA-SY-CCS Rev. 2, "Component Component Cooling System".

System.

[29] EA-PSA-PZR-SRV-1 2-03 Rev. 0, "Update EA-PSA-PZR-SRV-12-03 Update of Palisades PRA Consequential Pressurizer Methodology.

Safety Valve Operation Data and Methodology".

[30] Initiating Event Frequencies for Loss of Coolant EA-PSA-LOCA-IE-12-02 Rev. 0, "Initiating Assessment.

Accidents for the Palisades Nuclear Plant Probabilistic Risk Assessment".

[31] Level 1/Large Early Release Frequency ASME/ANS RA-S-2008a, Standard for Level1/Large Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME and the American Nuclear Society, December 2008.

[32] NEI 05-04, Revision 1 1 (Draft), "Process Process for Performing Follow-On PRA Peer Reviews Events), Nuclear Energy Institute, November Using the ASME PRA Standard (Internal Events),"

2007.

[33] SCIENTECH SCI Palisades Fire PRA Peer Review to Requirements in ENTECH document 17825-1, "Palisades Level 1/Large Early Release Frequency Part 4 of the ASME/ANS Standard for Level1/Large Applications.

Probabilistic Risk Assessments for Nuclear Power Plant Applications".

[34] EA-PSA-INTFLOOD-1 3-06 Vol. 1, "Palisades EA-PSA-INTFLOOD-13-06 Palisades Internal Flooding Analysis for Internal Components, Events PRA - Identification of Flood Areas, Flood Sources, and Impacted Components,"

Rev. 0, December, 2013.

[35] Palisades Internal Flooding Analysis for Internal EA-PSA-INTFLOOD-13-06 Vol. 2, "Palisades Events PSA - Initiating Event Frequencies for Flooding Events" Events Rev. 0, December December2013.

2013.

[36] EA-PSA-INTFLOOD-1 3-06 Vol. 3, "Palisades EA-PSA-INTFLOOD-13-06 Palisades Internal Flooding Analysis for Internal Frequency, Rev. 0, December Events PRA - Calculation of Core Damage Frequency,"

December2013.

2013.

[37] Pipe Rupture Frequencies for Internal Flooding PRAs, Electric Power Research Institute, "Pipe 3, Technical Report 3002000079, Final Report, April 2013.

Revision 3,"

[38] EA-PSA-FLOOD-IE-1 3-02 Rev. 0, "Internal EA-PSA-FLOOD-IE-13-02 Internal Flood Initiating Event Frequencies for the PRA.

Palisades PRA".

[39] EA-PSA-RI-ISI-00-INDIRECT ANALYSIS, RI-ISI Indirect Effects Evaluation, June 2000.

EA-PSA-RI-ISI-OO-INDIRECT

[40] AOP-39 Rev. 0, Palisades Abnormal Operating Procedure, "Internal Internal Plant Flooding".

Flooding.

[41] EN-NE-G-012 Rev. 0, Entergy Engineering Guide, "InternalInternal Flooding Analysis Guidelines.

Guidelines".

[42] EN-NE-G-012 Rev. 0, Entergy Engineering Guide, "InternalInternal Flooding Analysis Guidelines.

Guidelines".

[43] EA-PSA-INTFLOOD-1 3-06 Rev. 0, "Palisades EA-PSA-INTFLOOD-13-06 Palisades Flooding Analysis for Internal Events PSA, December PSA," December2013.

2013.

Page 55 of 55