Text

Proposed Alternative to Pvngs' ASME Section XI Inservice Inspection Program for ASME Code Category B-F, B-J, C-F-1, and C-F-2 Piping (Relief Request 32)
	ML060310511
Person / Time
Site:	Palo Verde
Issue date:	01/16/2006
From:	Mauldin D; Arizona Public Service Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	102-05398-CDM/SAB/RJR
	Download: ML060310511 (53)
	v • d • e

L AL ; 10 CFR 50.55a(a)(3)(i)(g)

David Mauldin Vice President Mail Station 7605 Palo Verde Nuclear Nuclear Engineering TEL (623) 393-5553 P.O. Box 52034 Generating Station and Support FAX (623) 393-6077 Phoenix, AZ 85072-2034 102-05398-CDM/SAB/RJR January 16, 2006 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Dear Sirs:

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2 and 3 Docket Nos. STN 50-528, STN 50-529, STN 50-530 Proposed Alternative to PVNGS' ASME Section Xi Inservice Inspection Program for ASME Code Category B-F, B-J, C-F-1, and C-F-2 Piping (Relief Request 32)

Pursuant to 10 CFR 50.55a(a)(3)(i), Arizona Public Service (APS) is proposing alternatives to 10 CFR 50.55a(g), "Inservice Inspection Requirements." APS is proposing to use the enclosed Risk-Informed Inservice Inspection Program as an alternative to the current inservice inspection program requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, 1992 Edition, 1992 Addenda, Section Xl, sub-articles IWB-2412 and IWC-2412 for the selection and examination of ASME Class 1 and 2 Code Category B-F, B-J, C-F-1, and C-F-2 piping welds. The proposed Palo Verde Nuclear Generating Station (PVNGS) alternative is based on the risk-informed process described in Electric Power Research Institute (EPRI) Technical Report TR-1 12657 Revision B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure," and the associated NRC Safety Evaluation. This program meets the intent and principles of NRC Regulatory Guides 1.174 and 1.178.

The enclosure to this letter contains APS' Relief Request 32. The attachment to the enclosure provides the proposed PVNGS RI-ISI program. Relief Request 32 supports the conclusion that the proposed alternative provides an acceptable level of quality and safety as required by 10 CFR 50.55a(a)(3)(i).

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance AzQ 7 Callaway

Comanche Peak

Diablo Canyon

Palo Verde

South Texas Project

Wolf Creek

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Relief Request 32 Page 2 Arizona Public Service requests NRC approval of the PVNGS RI-ISI Program by July 28, 2006. There are no new commitments being made in this request. If you have any questions, please contact Thomas N. Weber at (623) 393-5764.

Sincerely, CDM/SAB/RJR/ca

Enclosure:

Relief Request 32 - Palo Verde Nuclear Generating Station Units 1, 2 and 3.

Attachment:

APS' Risk-Informed Inservice Inspection Program Plan for Palo Verde Nuclear Generating Station, Revision 0.

cc: B. S. Malleft NRC Region IV Regional Administrator M. B. Fields NRC NRR Project Manager G. G. Warnick NRC Senior Resident Inspector for PVNGS

Enclosure Relief Request 32 Palo Verde Nuclear Generating Station Unit 1, 2 and 3

Relief Request 32 PVNGS Units 1, 2, and 3 Background Information Inservice inspections (ISI) are currently performed on Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3 piping in accordance with the requirements of the ASME Boiler and Pressure Vessel Code Section Xl, 1992 Edition, 1992 Addenda. All three units are currently in the second 10-year inspection interval as defined by the ASME Code for Program B. The start and end dates for each unit's second 10-year inspection interval is listed below.

Unit Start End 1 July 18,1998 July 17, 2008 2 March 18,1997 March 17, 2007 3 January 11,1998 January 10, 2008 The objective of this submittal is to request the use of a risk-informed (RI) process for the inservice inspection (ISI) of Class 1 and 2 piping. The proposed RI-ISI process used in this submittal is described in Electric Power Research Institute (EPRI) Technical Report TR-1 12657 Rev. B-A, "Revised Risk-informed Inservice Inspection Evaluation Procedure" (Reference 1). The RI-ISI application was also conducted in a manner consistent with ASME Code Case N-578, "Risk-informed Requirements for Class 1, 2, and 3 Piping, Method B." However, ASME Code Case N-578 is not the basis for the RI-ISI relief request. The RI-ISI application for PVNGS was conducted strictly in accordance with EPRI TR-1 12657.

I. ASME Code Component(s) Affected PVNGS Units: 1, 2 and 3 Code Class: 1 and 2

Description:

All Class 1 and 2 pressure retaining piping welds Categories: B-F, B-J, C-F-1, and C-F-2 Item Numbers: B5.40, B9.10, B9.20, B9.30 and B9.40 C5.10, C5.20, C5.30, C5.40, C5.50, C5.60, C5.70 and C5.80 II. Applicable Code Addition and Addenda The second 10-year inservice inspection interval code for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3 is the American Society of Mechanical Engineers (ASME) Code, Section Xl, 1992 Edition, 1992 Addenda.

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Relief Request 32 PVNGS Units 1, 2, and 3 IIl. Applicable Code Requirements ASME Section Xi, IBW-2500(a) states:

"Components shall be examined and tested as specified in Table IWB-2500-1.

The method of examination for the components and parts of the pressure retaining boundaries shall comply with those tabulated in Table IWB-2500-1 except where alternate examination methods are used that meet the requirements of IWA-2240."

Table IWB-2500-1, Examination Category, B-F, requires a volumetric and/or surface examination on all welds.

Table IWB-2500-1, Examination Category, B-J, requires a volumetric and/or surface examination on all welds.

ASME Section Xl, ICW-2500(a) states:

"Components shall be examined and tested as specified in Table IWC-2500-1.

The method of examination for the components and parts of the pressure retaining boundaries shall comply with those tabulated in Table IWC-2500-1 except where alternate examination methods are used that meet the requirements of IWA-2240."

Table IWC-2500-1, Examination Category, C-F-1, requires a volumetric and/or surface examination on all welds.

Table IWC-2500-1, Examination Category, C-F-2, requires a volumetric and/or surface examination on all welds.

IV. Reason For Request Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative would provide an acceptable level of quality and safety. In the October 28, 1999 Safety Evaluation from the NRC to EPRI (Reference 2), the NRC concluded that the risk informed inservice inspection program as described in EPRI TR-1 12657, Revision B, (now Revision B-A) is a sound technical approach and will provide an acceptable level of quality and safety pursuant to 10 CFR 50.55a for the proposed alternative to the piping inservice inspection requirements with regard to the number of locations, locations of inspections, and methods of inspection.

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Relief Request 32 PVNGS Units 1, 2, and 3 V. Proposed Alternative and Basis for Use The APS proposed alternative is described in the Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0, and is included as an attachment to this enclosure. APS has applied the methodology of the EPRI Technical Report TR-1 12657, Revision B-A (Reference 1) and associated NRC Safety Evaluation (SE) (Reference 2) in the development of the proposed PVNGS RI-ISI Program Plan with two deviations as described in Section 3 of the attachment. As described in the NRC SE associated with EPRI TR-112657 (Reference 2), the use of this methodology for the selection and subsequent examination of Class 1 and 2 piping welds provides an acceptable level of quality and safety.

Additionally, all piping components, regardless of risk classification, will continue to receive ASME Code-required pressure and leak testing, as part of the current ASME Section Xl program. Visual examinations (VT-2) are scheduled In accordance with the PVNGS pressure and leak test program, which remains unaffected by the proposed Rl-ISI program.

VI. Duration of Proposed Alternative Relief is requested for the third period of the second ten-year interval of the PVNGS Inservice Inspection Program.

VII. Conclusion 10 CFR 50.55a(a)(3) states:

"Proposed alternatives to the requirements of paragraphs (c), (d), (e), (f), (g), and (h) of this section or portions thereof may be used when authorized by the Director of the Office of Nuclear Reactor Regulation. The applicant shall demonstrate that:

(i) The proposed alternatives would provide an acceptable level of quality and safety, or (ii) Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety."

The proposed alternative RI-ISI program is consistent with EPRI TR-1 12657 and the associated NRC SE with deviations. As stated in the NRC SE, the Staff concluded that the proposed RI-ISI program, as described in EPRI TR-1 12657, Revision B, (now Revision B-A), is a sound technical approach and will provide an Page 3

Relief Request 32 PVNGS Units 1, 2, and 3 acceptable level of quality and safety pursuant to 10 CFR 50.55a. The APS proposed alternative, including deviations, discussed in the attachment to this request, provides an acceptable level of quality and safety. Therefore, APS requests that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).

APS requests approval of this relief by July 28, 2006.

VilI. References

1. Electric Power Institute Technical Report TR-1 12657, Revision B-A, "Revised Risk-Informed Inservice Evaluation Procedure," dated December 1999.

2. W. H. Bateman (U.S. NRC) to G.L. Vine (EPRI) letter dated October 28, 1999 transmitting "Safety Evaluation Report Related to EPRI Risk-Informed Inservice Inspection Evaluation Procedure (EPRI TR-112657, Revision B, July 1999)."

IX. Precedent San Onofre Nuclear Generating Station - Approved February 28, 2005 -

ML050940366 Cooper Nuclear Station - Approved December 9, 2004 - ML043440051 Hope Creek Generating Station -Approved December 8, 2004 - ML043080161 Farley Nuclear Power Plant - Approved March 9, 2004 - ML040700258 Page 4

Attachment APS' Risk-informed Inservice Inspection Program Plan for Palo Verde Nuclear Generating Station, Revision 0.

Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table of Contents

1. Introduction 1.1. Relation to NRC Regulatory Guides 1.174 and 1.178 1.2. PRA Quality

2. Proposed Alternative to Current Inservice Inspection Programs 2.1 ASME Section Xi 2.2 Augmented Programs

3. Risk-informed ISI Process 3.1 Scope of Program 3.2 Consequence Evaluation 3.3 Failure Potential Assessment 3.4 Risk Characterization 3.5 Element and NDE Selection 3.5.1 Additional Examinations 3.5.2 Program Relief Requests 3.6 Risk Impact Assessment 3.6.1 Quantitative Analysis 3.6.2 Defense-in-Depth

4. Implementation and Monitoring Program

5. Proposed Inservice Inspection Program Plan Change

6. References/Documentation

7. Abbreviations i

Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0

1. INTRODUCTION The Palo Verde Nuclear Generating Station (PVNGS) is currently in the second inservice inspection (ISI) interval as defined by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Section Xi Code for Inspection Program B. PVNGS plans to start implementing a risk-informed inservice inspection (RI-ISI) program during the third inspection period. The objective of this submittal is to request the use of a risk-informed (RI) process for the ISI of Class 1 and 2 piping.

The proposed RI-ISI process used in this submittal is described in Electric Power Research Institute (EPRI) Technical Report (TR) 112657 Rev. B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure" (Reference 6.1). The RI-ISI application was also conducted in a manner consistent with ASME Code Case N-578, "Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method B" (Reference 6.2). However, ASME Code Case N-578 is not the basis for the RI-ISI relief request. The RI-ISI application for PVNGS was conducted strictly in accordance with EPRI TR-1 12657.

1.1. Relation to NRC Regulatory Guides 1.174 and 1.178, RI As a RI application, this submittal meets the intent and principles of Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis" (Reference 6.3) and Regulatory Guide 1.178, Revision 1, "An Approach for Plant-Specific Risk-informed Decision making Inservice Inspection of Piping" (Reference 6.4). Further information is provided in Section 3.6.2 relative to defense-in-depth.

1.2. PRA Quality The initial PVNGS probabilistic risk assessment (PRA) model was described in APS' letter 161-04750, dated April 28, 1992, in response to Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities," and has been periodically updated. The RI-ISI analysis is documented in APS Engineering Study 13-NS-C097, "Risk-Informed Inservice Inspection Consequence Evaluation of Class 1 and 2 Piping," and was developed using APS Engineering Study 13-NS-C029, "Interim PRA Change Documentation,"

Revision 13. The PVNGS PRA model is dominated by a loss of either train of class power, fire and uncomplicated reactor trip. Total combined internal and fire Core Damage Frequency (CDF) is 1.64E-5/yr, and total combined Large Early Release Frequency (LERF) is 1.74E-6/yr.

PVNGS maintains a database to track potential update issues that may impact the PRA model or its documentation. Each current issue identified in the database was evaluated for potential impact on the RI-ISI analysis. The evaluation concluded that only one outstanding issue would impact the Page 1 of 44

Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 proposed RI-ISI analysis results. Impact No. 2004-179 addresses the incorrect assignment of a Steam Line Break Level 1 consequence sequence to a Level 2 event tree. The resolution of this impact was incorporated into the Revision 13 model and was used for the PVNGS RI-ISI analysis.

The PVNGS PRA has undergone numerous independent reviews. All found the PVNGS PRA model to be acceptable. A brief summary is provided below.

Independent Reviewer Scope CE Owner's Group Peer Assessment - Original Industry Peer Assessment of November 1999 the Internal Events at Power PVNGS PRA model.

ERIN Engineering and Research - A review of the resolution of the issues December 2000. identified in the CE Owners group Peer Assessment.

RELCON AB (Suppliers of the Risk A review of the application of the Risk Spectrum PRA Software) - August 2001. Spectrum PRA software to the PVNGS PRA model.

Potential weaknesses identified during these reviews are not considered significant and were either incorporated into the model or the tracking databases described above and were considered for this application.

The staffs review of the original APS IPE submittal documented in a letter dated July 1,1994, identified no areas for PRA improvement or weaknesses.

Based on this and the independent reviews subsequently performed, the current PRA model is judged to be of an acceptable quality to support this application.

2. PROPOSED ALTERNATIVE TO CURRENT ISI PROGRAMS 2.1. ASME Section Xl ASME Section Xi Examination Categories B-F, B-J, C-F-1 and C-F-2 currently contain the requirements for the nondestructive examination (NDE) of Class 1 and 2 piping components. An alternative RI-ISI Program for piping is described in EPRI TR-1 12657. The proposed PVNGS RI-ISI Program will be substituted for the current program for Class 1 and 2 piping (Examination Categories B-F, B-J, C-F-1 and C-F-2) in accordance with 10 CFR 50.55a(a)(3)(i) as it provides an acceptable level of quality and safety. Other non-related portions of the ASME Section Xl Code will be unaffected. EPRI TR- 12657 provides the requirements for defining the relationship between the RI-ISI Program and the remaining unaffected portions of ASME Section Xl.

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Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 2.2. Augmented Programs The plant augmented inspection programs listed below were considered during the RI-ISI application. It should be noted that this section documents only those plant augmented inspection programs that address common piping with the RI-ISI application scope (i.e., Class 1 and 2 piping).

A plant augmented inspection program has been implemented at PVNGS in response to NRC Bulletin 88-08, "Thermal Stresses in Piping Connected to Reactor Coolant Systems." The thermal fatigue concern addressed by this bulletin was explicitly considered in the application of the EPRI RI-ISI process and is subsumed by the RI-ISI Program.

A plant augmented inspection program has been implemented at PVNGS in response to NRC Bulletin 88-11, "Pressurizer Surge Line Stratification." The thermal fatigue concern addressed by this bulletin was explicitly considered in the application of the EPRI RI-ISI process and is subsumed by the RI-ISI Program.

A plant augmented inspection program has been implemented at PVNGS in response to NRC IE Bulletin 79-13, "Cracking in Feedwater System Piping."

This bulletin was a precursor to NRC Information Notice 93-20, "Thermal Fatigue Cracking of Feedwater Piping to Steam Generators." The thermal fatigue concern addressed by this bulletin and information notice was explicitly considered in the application of the EPRI RI-ISI process and is subsumed by the RI-ISI Program.

A plant augmented inspection program has been implemented at PVNGS in response to NRC Information Notice 97-19, "Safety Injection System Weld Flaw at Sequoyah Nuclear Power Plant, Unit 2." The intergranular stress corrosion cracking concern addressed by this information notice was explicitly considered in the application of the EPRI RI-ISI process. As a result, the RI-ISI Program effectively subsumes this plant augmented inspection program.

A plant augmented inspection program has been implemented at PVNGS in response to NRC IE Bulletin 79-17, "Pipe Cracks in Stagnant Borated Water Systems at PWR Plants." The intergranular stress corrosion cracking concern addressed by this bulletin was explicitly considered in the application of the EPRI RI-ISI process and is subsumed by the RI-ISI Program.

The plant augmented inspection program for high energy line breaks outside containment, implemented in response to Updated Final Safety Analysis Report (UFSAR) Section 6.6.8, "Augmented Inservice Inspection to Protect against Postulated Piping Failures," is not affected or changed by the RI-ISI Program. It should be noted, however, that this piping is the Page 3 of 44

Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 subject of a separate and independent assessment being conducted in accordance with EPRI Report 1006937 Rev. 0-A, "Extension of the EPRI Risk-Informed Inservice Inspection (RI-ISI) Methodology to Break Exclusion Region (BER) Programs," dated August 2002.

The plant augmented inspection program for flow accelerated corrosion (FAC) per NRC Generic Letter 89-08 is relied upon to manage this damage mechanism but is not otherwise affected or changed by the RI-ISI Program.

3. RISK-INFORMED ISI PROCESS The process used to develop the proposed PVNGS RI-ISI Program conformes to the methodology described in EPRI TR-1 12657 and consisted of the following steps:

Scope Definition

Consequence Evaluation

Failure Potential Assessment

Risk Characterization

Element and NDE Selection

Risk Impact Assessment

Implementation Program

Feedback Loop A deviation to the EPRI RI-ISI methodology has been implemented in the Failure Potential Assessment step. Table 3-16 of EPRI TR-1 12657 contains criteria for assessing the potential for thermal stratification, cycling and striping (TASCS). Key attributes for horizontal or slightly sloped piping greater than 1" nominal pipe size (NPS) include:

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

2. 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. Potential exists for convective heating in dead-ended pipe sections connected to a source of hot fluid, or

4. Potential exists for two phase (steam/water) flow, or

5. Potential exists for turbulent penetration into a relatively colder branch pipe connected to header piping containing hot fluid with turbulent flow, AND Page 4 of 44

Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0

> AT > 500 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 actual AT assumed equal to the greatest potential AT for the transient, will identify locations where stratification is likely to occur, but allows for no assessment of severity. As such, many locations will be identified as subject to TASCS where no significant potential for thermal fatigue exists. The critical attribute missing from the existing methodology that would allow consideration of fatigue severity is a criterion that addresses the potential for fluid cycling.

The methodology used for assessing TASCS at PVNGS is identical to that in EPRI Technical Report 1000701, "Thermal Fatigue Management Guideline (MRP-24),"

dated January 2001, for the specific lines covered by guideline. In addition, the underlying basis of MRP-24 was used for assessing TASCS for other lines. Thus, the methodology for assessing TASCS is fully consistent with MRP-24. The impact of this additional consideration on the existing TASCS susceptibility criteria is presented below.

> Turbulent penetration TASCS Turbulent penetration typically occurs in lines connected to piping containing hot flowing fluid. In the case of downward sloping lines that then turn horizontal, significant top-to-bottom cyclic ATs can develop in the horizontal sections ifthe horizontal section is less than about 25 pipe diameters from the reactor coolant piping. Therefore, TASCS is considered for this configuration.

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 keep the line filled with hot water. Ifthere is no potential for in-leakage towards the hot fluid source from the outboard end of the line, this will result in a well-mixed fluid condition where significant top-to-bottom ATs will not occur. Therefore TASCS is not considered for these configurations. Even in fairly long lines, where some heat loss from the outside of the piping will tend to occur and some fluid stratification may be present, there is no significant potential for cycling as has been observed for the in-leakage case. The effect of TASCS will not be significant under these conditions and can be neglected.

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Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0

> Low flow TASCS In some situations, the transient startup of a system (e.g., RHR suction piping) creates the potential for fluid stratification as flow is established. In cases where no cold fluid source exists, the hot flowing fluid will fairly rapidly displace the cold fluid in stagnant lines, while fluid mixing will occur in the piping further removed from the hot source and stratified conditions will exist only briefly as the line fills with hot fluid. As such, since the situation is transient in nature, it can be assumed that the criteria for thermal transients (TT) will govern.

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

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

In summary, these additional considerations for determining the potential for thermal fatigue as a result of the effects of TASCS provide an allowance for the consideration of cycle severity. The above criteria have previously been submitted by EPRI for generic approval in letters dated February 28, 2001 and March 28, 2001, from P.J. O'Regan (EPRI) to Dr. B. Sheron (USNRC), "Extension of Risk-Informed Inservice Inspection Methodology" (ADAMS accession numbers ML010650169 and ML011070238 respectively). The methodology used in the PVNGS RI-ISI application for assessing TASCS potential conforms to these updated criteria. Final materials reliability program (MRP) guidance on the subject of TASCS will be incorporated into the PVNGS RI-ISI application if warranted.

A second deviation to the EPRI RI-ISI methodology has been implemented in the Element and NDE Selection step. A 10% selection sampling is required of risk category 4 inspection locations for each system in the RI-ISI application scope. In the charging system (CH) for all three units, an additional risk category 2 inspection location was selected for examination instead of making a risk category 4 selection as indicated in the table below.

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Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Unit Risk DMs Weld RI-ISI Selections category C Count Required l Made 2 TASCS,TT 2 6 2 1 TT 22 5 4 None 5 1 0 2 TASCS,TT 2 6 2 2 Yr 22 __ _ _ _ _5 4 None 5 1 0 2 TASCS, TT 2 6 2 3 None 22 5 4 None 5 1 0 The five risk category 4 CH welds per unit are located just beyond the affected regions of the charging lines that are potentially subjected to thermal transients when flow is restored after a loss of charging transient. PVNGS conservatively elected to choose an additional risk category 2 inspection location for examination from the region subject to thermal transients, in lieu of making a risk category 4 selection in the unaffected piping section.

3.1. Scope of Program The systems included in the proposed PVNGS RI-ISI Program are provided in Tables 3.1-1, 3.1-2 and 3.1-3 for Units 1, 2 and 3, respectively. The piping and instrumentation diagrams and additional plant information including the existing plant ISI Program were used to define the Class 1 and 2 piping system boundaries.

3.2. Consequence Evaluation The consequence(s) of pressure boundary failures were evaluated and ranked based on their impact on core damage and containment performance (i.e.,

isolation, bypass and large early release). The consequence evaluation included an assessment of shutdown and external events. The impact on these measures due to both direct and indirect effects was considered using the guidance provided in EPRI TR-1 12657.

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Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 3.3. 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 EPRI TR-1 12657, with the exception of the previously stated deviation.

Tables 3.3-1, 3.3-2 and 3.3-3 summarize the failure potential assessment by system for each degradation mechanism that was identified as potentially operative for Units 1, 2 and 3, respectively.

3.4. Risk Characterization In the preceding steps, each run of piping within the scope of the program was evaluated to determine its impact on core damage and containment performance (i.e., isolation, bypass and large, early release) as well as its potential for failure. Given the results of these steps, piping segments are then defined as continuous runs of piping potentially susceptible to the same type(s) of degradation and whose failure will result in similar consequence(s).

Segments are then ranked based upon their risk significance as defined in EPRI TR-112657.

The results of these calculations are presented in Tables 3.4-1, 3.4-2 and 3.4-3 for Units 1, 2 and 3, respectively.

3.5. Element and NDE Selection In general, EPRI TR-1 12657 requires that 25% of the locations in the high risk region and 10% of the locations in the medium risk region be selected for inspection using appropriate NDE methods tailored to the applicable degradation mechanism. In addition, per Section 3.6.4.2 of EPRI TR-1 12657, if the percentage of Class 1 piping locations selected for examination falls substantially below 10%, then the basis for selection needs to be investigated.

For the PVNGS, the percentage of Class 1 piping welds selected strictly for RI-ISI purposes exceeds 10% for all three units. It should be noted that this sampling percentage for Class I piping locations includes both socket and non-socket welds. If only non-socket welded locations are considered, the percentage of Class 1 piping welds selected for examination would be even greater.

The above sampling percentage does not take credit for any inspection locations selected for examination per the plant's augmented inspection program for FAC beyond those selected per the RH-ISI process. It should be noted that no FAC examinations are being credited to satisfy RI-ISI selection Page 8 of 44

Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 requirements. Inspection locations selected for RI-ISI purposes that are in the FAC Program will be subjected to an independent examination to satisfy the RI-ISI Program requirements.

A brief summary is provided below, and the results of the selections are presented in Tables 3.5-1, 3.5-2 and 3.5-3 for Units 1, 2 and 3, respectively.

Section 4 of EPRI TR-1 12657 provides guidance for determining the examination requirements.

Unit Class I Piping Welds(1) Class 2 Piping Welds(2)l All Piping Welds(3) i Total l_Selected Total Selected Total Selected 1 596 61 J 2192 42 2788 103 2 l 573 61 l 2221 47 2794 108 l 3 1 570 61 [ 2172 41 2742 102 Notes

1. Includes all Category B-F and B-J locations.

2. Includes all Category C-F-1 and C-F-2 locations.

3. All in-scope piping components, regardless of risk classification, will continue to receive Code required pressure testing, as part of the current ASME Section Xl Program. VT-2 visual examinations are scheduled in accordance with the station's pressure test program that remains unaffected by the RI-ISI Program.

3.5.1. Additional Examinations The RI-ISI Program in all cases will determine through an engineering evaluation the root cause of any unacceptable flaw or relevant condition found during examination. The evaluation will include the applicable service conditions and degradation mechanisms to establish that the element(s) will still perform their intended safety function during subsequent operation. Elements not meeting this requirement will be repaired or replaced.

The evaluation will include whether other elements in the segment or additional segments are subject to the same root cause conditions.

Additional examinations will be performed on those elements with the same root cause conditions or degradation mechanisms. The additional examinations will include high risk significant elements and medium risk significant elements, if needed, up to a number equivalent to the number of elements required to be inspected on the segment or segments during the current outage. If unacceptable flaws or relevant conditions are again found similar to the initial problem, the remaining elements identified as susceptible will be examined during the current outage. No additional Page 9 of 44

Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 examinations will be performed if there are no additional elements identified as being susceptible to the same root cause conditions.

3.5.2. Program Relief Requests An attempt has been made to select RI-ISI locations for examination such that a minimum of >90% coverage (i.e., Code Case N-460 criteria) is attainable. However, some limitations will not be known until the examination is performed, since some locations may be examined for the first time by the specified techniques. In instances where locations are found at the time of the examination that do not meet the >90% coverage requirement, the process outlined in EPRI TR-1 12657 will be followed.

None of the existing PVNGS relief requests are being withdrawn due to the RI-ISI application.

3.6. Risk Impact Assessment The RI-ISI Program has been conducted in accordance with Regulatory Guide 1.174 and the requirements of EPRI TR-1 12657, 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 identified the allocation of segments into High, Medium, and Low risk regions of the EPRI TR-1 12657 and ASME Code Case N-578 risk ranking matrix, and then determined for each of these risk classes what inspection changes are proposed for each of the locations in each segment.

The changes include changing the number and location of inspections within the segment and in many cases improving the effectiveness of the inspection to account for the findings of the RI-ISI degradation mechanism assessment. For example, for locations subject to thermal fatigue, examinations will be conducted on an expanded volume and will be focused to enhance the probability of detection (POD) during the inspection process.

3.6.1. Quantitative Analysis Limits are imposed by the EPRI methodology to ensure that the change in risk of implementing the RI-ISI Program meets the requirements of Regulatory Guides 1.174 and 1.178. The EPRI criterion requires that the cumulative change in core damage frequency (CDF) and large early release frequency (LERF) be less than 1E-07 and 1E-08 per year per system, respectively.

PVNGS conducted a risk impact analysis using the requirements of Section 3.7 of EPRI TR-1 12657. The analysis estimates the net change in Page 10 of 44

Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 risk due to the positive and negative influence of adding and removing locations from the inspection program. A risk quantification was performed using the "Simplified Risk Quantification Method" described in Section 3.7 of EPRI TR-1 12657. The conditional core damage probability (CCDP) and conditional large early release probability (CLERP) used for high consequence category segments was based on the highest evaluated CCDP (1.4E-02) and CLERP (1.4E-03), whereas, for medium consequence category segments, bounding estimates of CCDP (1E-04) and CLERP (1E-05) were used. The likelihood of pressure boundary failure (PBF) is determined by the presence of different degradation mechanisms and the rank is based on the relative failure probability. The basic likelihood of PBF for a piping location with no degradation mechanism present is given as x0 and is expected to have a value less than 1E-08. Piping locations identified as medium failure potential have a likelihood of 20x 0. These PBF likelihoods are consistent with References 9 and 14 of EPRI TR-1 12657. 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 RI-ISI approach.

Tables 3.6-1, 3.6-2 and 3.6-3 present summaries of the RI-ISI Program versus 1992 ASME Section Xl Code Edition and Addenda program requirements and identifies on a per system basis each applicable risk category for Units 1, 2 and 3, respectively. The presence of FAC was adjusted for in the performance of the quantitative analysis by excluding its impact on the risk ranking. The exclusion of the impact of FAC on the risk ranking and therefore in the determination of the change in risk is performed, because FAC is a damage mechanism managed by a separate, independent plant augmented inspection program. The RI-ISI Program credits and relies upon this plant augmented inspection program to manage this damage mechanism. The plant FAC Program will continue to determine where and when examinations shall be performed. Hence, since the number of FAC examination locations remains the same "before" and "after" and no delta exist, there is no need to include the impact of FAC in the performance of the risk impact analysis. However, in an effort to be as informative as possible, for those systems where FAC is present, Tables 3.6-1, 3.6-2 and 3.6-3 present the information in such a manner as to depict what the resultant risk categorization is both with and without consideration of FAC. This is accomplished by enclosing the FAC damage mechanism, as well as all other resultant corresponding changes (failure potential rank, risk category and risk rank), in parentheses. Again, this has only been done for information purposes, and has no impact on the assessment itself. The use of this approach to depict the impact of degradation mechanisms managed by plant augmented inspection programs on the risk categorization is consistent with that used in the Page 11 of 44

Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 delta risk assessment for the Arkansas Nuclear One, Unit 2 (ANO-2) pilot application. An example is provided below.

Risk Consequence Failure Potential System Ckgory) Rank DMs Rank Cteor In this example if FAC is not considered, the failure potential rank is 'medium' instead of 'high" based on the TASCS and TT damage mechanisms. When a 'medium" failure potential rank is combined with a "medium" consequence rank, it results in risk category 5 (medium" risk) being assigned instead of risk category 3 ("high" risk).

SG 5 (3) Medium (High) Medium TASCS, TT, (FAC)' Medium (f In this example if FAC were considered, the failure potential rank would be "high" instead of "medium." If a "high" failure potential rank were combined with a "medium" consequence 1 rank, it would result in risk category 3 ("high" risk) being assigned instead of risk category 5 ("medium"risk).

Note

1. The risk rank is not included in Tables 3.6-1, 3.6-2 or 3.6-3 but it is included in Tables 5-2-1, 5-2-2 and 5-2-3.

As indicated in the following tables, this evaluation has demonstrated that unacceptable risk impacts will not occur from implementation of the PVNGS Rl-ISI Program, and satisfies the acceptance criteria of Regulatory Guide 1.174 and EPRI TR-112657.

Unit I Risk Impact Results System( P ARiskCDF ARiskLERF W/ POD i w/o POD w/ POD l w/o POD l RC -5.46E-09 1.02E-08 -5.46E-10 1.02E-09 CH -1.76E-08 -9.80E-09 -1.76E-09 -9.80E-10 Si 4.84E-10 5.OOE-10 4.84E-11 5.OOE-11 AF 4.60E-1I 7.OOE-11 4.60E-12 7.OOE-12 SG -2.64E-08 -1.02E-08 -2.64E-09 -1.02E-09 Total -4.90E-08 -9.17E-09 -4.90E-09 -9.17E-10 Note

1. Systems are described in Table 3.1-1.

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Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Unit 2 Risk Impact Results System IP ARiSkCDF ARiskLERF w/ POD l w/o POD w/ POD l who POD RC -5.46E-09 1.02E-08 -5.46E-10 1.02E-09 CH -1.76E-08 -9.80E-09 -1.76E-09 -9.80E-10 Si 4.74E-10 4.90E-10 4.74E-11 4.90E-11 AF 1.16E-10 1.40E-10 1.16E-11 1.40E-11 SG -2.39E-08 -5.96E-09 -2.39E-09 -5.96E-10 Total -4.64E-08 -4.91 E-09 -4.64E-09 -4.911E-10 Note

1. Systems are described in Table 3.1-2.

Unit 3 Risk Impact Results System( ) ARiSkCDF ARiSkLERF wi POD w/o POD w/ POD wlo POD RC -5.46E-09 1.02E-08 -5.46E-10 1.02E-09 CH -1.76E-08 -9.80E-09 -1.76E-09 -9.80E-10 Si 4.24E-10 4.40E-10 4.24E-11 4.40E-11 AF 1.16E-10 1.40E-10 1.16E-11 1.40E-11 SG -2.39E-08 -8.70E-09 -2.39E-09 -8.70E-10 Total -4.64E-08 -7.70E-09 -4.64E-09 -7.70E-10 Note

1. Systems are described in Table 3.1-3.

3.6.2. Defense-in-Depth The intent of the inspections mandated by ASME Section Xl for piping welds is to identify conditions such as flaws or indications that may be precursors to leaks or ruptures in a system's pressure boundary.

Currently, the process for picking inspection locations is based upon structural discontinuity and stress analysis results. As depicted in ASME White Paper 92-01-01 Rev. 1, "Evaluation of Inservice Inspection Requirements for Class 1, Category B-J Pressure Retaining Welds," this method has been ineffective in identifying leaks or failures. EPRI TR-112657 and Code Case N-578 provide a more robust selection process founded on actual service experience with nuclear plant piping failure data.

This process has two key independent ingredients: (1) a determination of each location's susceptibility to degradation, and (2) an independent assessment of the consequence of the piping failure. These two Page 13 of 44

Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 ingredients assure defense in depth is maintained. First, by evaluating a location's susceptibility to degradation, the likelihood of finding flaws or indications that may be precursors to leak or ruptures is increased.

Secondly, the consequence assessment effort has a single failure criterion. As such, no matter how unlikely a failure scenario is, it is ranked High in the consequence assessment, and at worst Medium in the risk assessment (i.e., Risk Category 4), if as a result of the failure there is no mitigative equipment available to respond to the event. In addition, the consequence assessment takes into account equipment reliability, and less credit is given to less reliable equipment.

All locations within the Class 1 and 2 pressure boundaries will continue to receive a system pressure test and visual VT-2 examination as currently required by the Code regardless of its risk classification.

4. IMPLEMENTATION AND MONITORING PROGRAM Upon approval of the PVNGS RI-ISI Program, procedures that comply with the guidelines described in EPRI TR-1 12657 will be prepared to implement and monitor the program. The new program will be integrated into the second inservice inspection interval for all three units. No changes to the Technical Specifications or Updated Final Safety Analysis Report are necessary for program implementation.

The applicable aspects of the ASME Code not affected by this change will be retained, such as inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements.

Existing ASME Section Xl program implementing procedures will be retained and modified to address the RI-ISI process, as appropriate.

The monitoring and corrective action program will contain the following elements:

A. Identify B. Characterize C. (1) Evaluate, determine the cause and extent of the condition identified (2) Evaluate, develop a corrective action plan or plans D. Decide E. Implement F. Monitor G. Trend The PVNGS RI-ISI Program is a living program requiring feedback of new relevant information to ensure the appropriate identification of high safety significant piping locations. As a minimum, risk ranking of piping segments will be reviewed and Page 14 of 44

Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 adjusted on an ASME period basis. In addition, significant changes may require more frequent adjustment as directed by NRC Bulletin or Generic Letter requirements, or by industry and plant specific feedback.

5. PROPOSED ISI PROGRAM PLAN CHANGE A comparison between the PVNGS RI-ISI Program and ASME Section Xl 1992 Code Edition and Addenda program requirements for in-scope piping is provided in Tables 5-1-1 and 5-2-1 for Unit 1, Tables 5-1-2 and 5-2-2 for Unit 2 and Tables 5-1-3 and 5-2-3 for Unit 3. Tables 5-1-1, 5-1-2 and 5-1-3 provide summary comparisons by risk region. Table 5-2-1, 5-2-2 and 5-2-3 provide the same comparison information, but in a more detailed manner by risk category, similar to the format used in Tables 3.6-1, 3.6-2 and 3.6-3.

Upon approval, APS plans to start implementing the PVNGS RI-ISI Program during the third inspection period. Once implemented, inspection locations selected per the RI-ISI process will replace those formerly selected per ASME Section Xl criteria.

The table below indicates the percentage of piping weld examinations required by ASME Section Xl that are expected to be completed in the second ISI interval for Examination Categories B-F, B-J, C-F-1 and C-F-2 and the percentage of inspection locations selected for examination using the PVNGS RI-ISI program process that will be examined in the third period of the current ten-year ISI interval.

Examination Percentages Expected at Unit the End of the 2 "dInterval ASME Section XI RI-ISI Program 1 63% 37%

2 61% 39%

58% 42%

Subsequent ISI intervals will implement 100% of the inspection locations selected for examination per the RI-ISI Program. Examinations shall be performed such that the period percentage requirements of ASME Section Xl, paragraphs IWB-2412 and IWC-2412 are met.

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Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0

6. REFERENCESIDOCUMENTATION 6.1. EPRI TR-1 12657, "Revised Risk-Informed Inservice Inspection Evaluation Procedure," Rev. B-A 6.2.ASME Code Case N-578, "Risk-informed Requirements for Class 1, 2, and 3 Piping, Method B, Section Xl, Division 1" 6.3. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-informed Decisions on Plant-Specific Changes to the Licensing Basis" 6.4. Regulatory Guide 1.178, Revision 1, "An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping" Supporting Onsite Documentation 13-NS-C067, "Risk-Informed Inservice Inspection Consequence Evaluation of Class 1 and 2 Piping for Palo Verde Nuclear Generating Station, Units 1, 2 and 3,"

Revision 1 USA-03-301, "Degradation Mechanism Evaluation for the Class 1, Class 2 and BER Program Piping Welds for Palo Verde Units 1, 2 and 3," Revision 1 ITS Document No. APS02.GO1, "Palo Verde Nuclear Generating Station, Units 1, 2 and 3, Risk-Informed Inservice Inspection Service History Review," Revision 0 USA-03-302, "Risk Ranking for the Palo Verde Nuclear Generating Station,"

Revision 0 USA-03-303, "Minutes of the Element Selection Meeting for the RI-ISI/RI-BER Project at Palo Verde Nuclear Generating Station," Revision 0 USA-03-304, "Risk Impact for the Palo Verde Nuclear Generating Station,"

Revision 0

7. ABBREVIATIONS ASME American Society of Mechanical Engineers BER Break Exclusion Region CC Crevice Corrosion CCDP Conditional Core Damage Probability CDF Core Damage Frequency Page 16 of 44

Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 CLERP Conditional Large Early Release Probability DM Degradation Mechanism ECSCC External Chloride Stress Corrosion Cracking EPRI Electric Power Research Institute E-C Erosion-Cavitation FAC Flow Accelerated Corrosion IGSCC Intergranular Stress Corrosion Cracking LERF Large Early Release Frequency MIC Microbiological Influenced Corrosion MRP Materials Reliability Program NDE Nondestructive Examination NRC Nuclear Regulatory Commission PIT Pitting POD Probability of Detection PRA Problistic Risk Assessment PVNGS Palo Verde Nuclear Generating Station PWSCC Primary Water Stress Corrosion Cracking RHR Residual Heat Removal RI-ISI Risk-informed Inservice Inspection SER Safety Evaluation Report SUR Surface SXI Section Xl TASCS Thermal Stratification, Cycling and Stripping TGSCC Transgranular Stress Corrosion Cracking TR Technical Report TT Thermal Transients VOL Volumetric Page 17 of 44

Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.1-1 Unit 1 - System Selection and Segment / Element Definition System Description Number of Segments [ Number of Elements RC - Reactor Coolant System(') 59 346 CH - Charging System (2) 22 164 SI - Safety Injection System(3) 167 1648 AF - Auxiliary Feedwater System(4) 11 59 SG - Steam Generator System(5) 65 571 Totals 324 2788 Notes

1. The system description includes the main loop hot and cold legs, shutdown cooling suction lines, pressurizer surge line, pressurizer main spray lines, pressurizer safeties, letdown line and drain lines.

2. The system description includes the charging line, letdown line, pressurizer auxiliary spray line and refueling water storage tank piping.

3. The system description includes high pressure safety injection piping, low pressure safety injection piping, shutdown cooling piping, common safety injection piping and containment spray piping.

4. The system description includes the auxiliary feedwater piping to the steam generators.

5. The system description includes main steam piping, feedwater piping, atmospheric dump piping, main steam to auxiliary feedwater turbine piping, blowdown piping and downcomer piping.

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i Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.1-2 Unit 2 - System Selection and Segment / Element Definition System Description Number of Segments Number of Elements RC - Reactor Coolant System(1) 60 338 CH - Charging System(2) 22 159 SI - Safety Injection System(3) 168 1623 AF - Auxiliary Feedwater System(4) 11 62 SG - Steam Generator System(5) 70 612 Totals 331 2794 Notes

1. The system description includes the main loop hot and cold legs, shutdown cooling suction lines, pressurizer surge line, pressurizer main spray lines, pressurizer safeties, letdown line and drain lines.

2. The system description includes the charging line, letdown line, pressurizer auxiliary spray line and refueling water storage tank piping.

3. The system description includes high pressure safety injection piping, low pressure safety injection piping, shutdown cooling piping, common safety injection piping and containment spray piping.

4. The system description includes the auxiliary feedwater piping to the steam generators.

5. The system description includes main steam piping, feedwater piping, atmospheric dump piping, main steam to auxiliary feedwater turbine piping, blowdown piping and downcomer piping.

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Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.1-3 Unit 3 - System Selection and Segment / Element Definition System Description Number of Segments Number of Elements 1

RC - Reactor Coolant System( ) 60 339 CH - Charging System(2) 22 162 SI - Safety Injection System(3) 168 1632 AF - Auxiliary Feedwater System(4) 12 58 SG - Steam Generator System(5) 61 551 Totals 323 2742 Notes

1. The system description includes the main loop hot and cold legs, shutdown cooling suction lines, pressurizer surge line, pressurizer main spray lines, pressurizer safeties, letdown line and drain lines.

2. The system description includes the charging line, letdown line, pressurizer auxiliary spray line and refueling water storage tank piping.

3. The system description includes high pressure safety injection piping, low pressure safety injection piping, shutdown cooling piping, common safety injection piping and containment spray piping.

4. The system description includes the auxiliary feedwater piping to the steam generators.

5. The system description includes main steam piping, feedwater piping, atmospheric dump piping, main steam to auxiliary feedwater turbine piping, blowdown piping and downcomer piping.

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Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.3-1 Unit 1 - Failure Potential Assessment Summary System(1) Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive TASCS TT IGSCC TGSCC l ECSCC PWSCC MIC PIT CC E-C FAC RC X X X CH X X_ _

AF X x SG X X x Note

1. Systems are described in Table 3.1-1.

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i Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.3-2 Unit 2 - Failure Potential Assessment Summary Syste Theral Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive TASCS lTT IGSCC TGSCC ECSCC PWSCC MIC_ PITccEC FAC RC X X Xl CH xx X X0:

AF X .X SX X X Note

1. Systems are described in Table 3.1-2.

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i Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.3-3 Unit 3 - Failure Potential Assessment Summary System( 1 ) Thermal Fatigue Stress Corrosion Cracking Localized Corrosion f Flow Sensitive TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC RC X X X CH X XX AF X x SG X X X Note

1. Systems are described in Table 3.1-3.

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a Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.4-1 Unit I - Number of Segments by Risk Category With and Without Impact of FAC

_ High Risk Region _ Medium Risk Region Low Risk Region System(r) Category I Category 2 Category 3 Category 4 Category 5 Category 6 Category 7 With [ Without With Without With J Without With Without ithout With Without With I Without RC 22 22 33 33 l _ 4 4 CH 2 2 1 1 19 19 Si 12 12 8 8 145 145 2 2 AF 1(2) 0 6 6 2 3 2 2 SG 11(3) 0 2 9 10(4) 0 10 14 8 14 24 28 Total 11 0 26 33 11 0 62 66 18 25 190 194 6 6 Notes

1. Systems are described in Table 3.1-1.

2. This segment becomes Category 5 after FAC is removed from consideration due to the presence of other "medium" failure potential damage mechanisms.

3. Of these eleven segments, seven become Category 2 after FAC is removed from consideration due to the presence of other "medium' failure potential damage mechanisms, and four become Category 4 after FAC is removed from consideration due to no other damage mechanisms being present.

4. Of these ten segments, six become Category 5 after FAC is removed from consideration due to the presence of other "medium" failure potential damage mechanisms, and four become Category 6 after FAC is removed from consideration due to no other damage mechanisms being present.

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Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.4-2 Unit 2 - Number of Segments by Risk Category With and Without Impact of FAC High Risk Region Medium Risk Region Low Risk Region System() Category I Category 2 Category 3 Category 4 Category 5 Category 6 Category 7 With Without With Without With Without With Without With Without With Without With Without RC 22 22 33 33 5 5 CH 2 2 1 1 19 19 Si 12 12 8 8 146 146 2 2 AF 1(2) 0 6 6 2 3 2 2 SG 0 6 17 00 9 13 8 14 24 26 Total 15 0 30 41 9 0 61 65 18 25 191 193 7 7 Notes

1. Systems are described in Table 3.1-2.

2. This segment becomes Category 5 after FAC is removed from consideration due to the presence of other 'medium" failure potential damage mechanisms.

3. Of these fifteen segments, eleven become Category 2 after FAC is removed from consideration due to the presence of other 'medium' failure potential damage mechanisms, and four become Category 4 after FAC is removed from consideration due to no other damage mechanisms being present.

4. Of these eight segments, six become Category 5 after FAC is removed from consideration due to the presence of other 'medium' failure potential damage mechanisms, and two become Category 6 after FAC is removed from consideration due to no other damage mechanisms being present.

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Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.4-3 Unit 3 - Number of Segments by Risk Category With and Without Impact of FAC High Risk Region Medium Risk Region Low Risk Region System(t) Category I Category 2 Category 3 Category 4 Category 5 Category 6 Category 7 With Without With Without With Without With Without With Without With Without With Without RC 22 22 33 33 5 5 CH 2 2 1 1 19 19 Si 12 12 8 8 146 146 2 2 AF 2(2) 0 6 6 2 4 2 2 G 2(3) 0 2 8 8'4' 0 10 16 7 11 22 26 Total 12 0 26 32 10 0 62 68 17 23 189 193 7 7 Notes

1. Systems are described in Table 3.1-3.

2. These two segments become Category 5 after FAC is removed from consideration due to the presence of other 'medium' failure potential damage mechanisms.

3. Of these twelve segments, six become Category 2 after FAC is removed from consideration due to the presence of other 'medium' failure potential damage mechanisms, and six become Category 4 after FAC is removed from consideration due to no other damage mechanisms being present.

4. Of these eight segments, four become Category 5 after FAC is removed from consideration due to the presence of other 'medium' failure potential damage mechanisms, and four become Category 6 after FAC is removed from consideration due to no other damage mechanisms being present.

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Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.5-1 Unit I - Number of Elements Selected for Inspection by Risk Category Excluding Impact of FAC High Risk Region Medium Risk Region Low Risk Region System(1) Category 1 Category 2 Category 3 Category 4 l Category 5 Category 6 l Category 7 Total lSelected Total Selected Total Selected Total JSelected Total Selected Total [Selected l Total J Selected RC 56 16 282 29 8 0 CH 24 7 5 0 135 0 Si 80 8 18 2 1544 0 6 0 AF 38 4 10 2 11 0 SG 45 12 174 19 32 4 320 0 Total 125 35 579 60 60 8 2010 0 14 0 Notes

1. Systems are described in Table 3.1-1.

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Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.5-2 Unit 2 - Number of Elements Selected for Inspection by Risk Category Excluding Impact of FAC High Risk Region Medium Risk Region Low Risk Region Total l Selected Total (Selected Total [Selected Total J Selected Total Selected Total l Selected Total l Selected RC 58 16 270 29 10 0 CH 24 7 5 0 130 0 Si __ 78 8 18 2 1521 0 6 0 AF 40 4 11 2 11 0 SG 44 12 222 23 42 5 304 0 Total 126 35 615 64 71 9 1966 0 16 0 Notes

1. Systems are described in Table 3.1-2.

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i Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.5-3 Unit 3 - Number of Elements Selected for Inspection by Risk Category Excluding Impact of FAC High Risk Region Medium Risk Region Low Risk Region T Category SysteSe() Categorye 2 Categ ry 3 Category 4 Category 5 Category 6 Category 7 Total Selected Total Selected Total Selected Total Selected Total Selected Total Selected Total Selected RC 58 16 271 29 10 0 CH 24 7 5 0 133 0 Si 74 8 18 2 1534 0 6 0 AF 38 4 10 2 10 0 SG 44 11 160 17 49 6 298 0 Total 126 34 548 58 77 10 1975 0 16 0 Notes

1. Systems are described in Table 3.1-3.

Page 29 of 44

i Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.6-1 Unit 1 - Risk Impact Analysis Results System( 1' I Category Consequence Failure Potential l Inspections l CDF Impact() LERF Impact(3 )

Rank DMs j Rank SXI(2) l RI-ISI Delta w/ POD w/o POD w/ POD w/o POD RC 2 High TASCS, TT Medium 4 2 -2 -1.68E-09 2.80E-09 -1.68E-10 2.80E-10 RC 2 High TT, PWSCC Medium 3 3 0 no change no change no change no change RC 2 High TASCS Medium 5 3 -2 -3.36E-09 2.80E-09 -3.36E-10 2.80E-10 RC 2 High TT Medium 5 2 -3 -8.40E-10 4.20E-09 -8.40E-11 4.20E-10 RC 2 High PWSCC Medium 6 6 0 no change no change no change no change RC 4 High None Low 35 29 -6 4.20E-10 4.20E-10 4.20E-11 4.20E-11 RC 7a Low None Low 0 0 0 no change no change no change no change RC Total -5.46E-09 1.02E-08 -5.46E-10 1.02E-09 CH 2 High TASCS, TT Medium 0 2 2 -5.04E-09 -2.80E-09 -5.04E-10 -2.80E-10 CH 2 High TT Medium 0 5 5 -1.26E-08 -7.OOE-09 -1.26E-09 -7.OOE-10 CH 4 High None Low 0 0 0 no change no change no change no change CH 6a Medium None Low 8 0 -8 negligible negligible negligible negligible CH Total -1.76E-08 -9.80E-09 -1.76E-09 -9.80E-10 Si 4 High None Low 15 8 -7 4.90E-10 4.90E-10 4.90E-11 4.90E-11 Si 5a Medium TASCS Medium 2 1 -1 -6.OOE-12 1.OOE-11 -6.00E-13 1.00E-12 Si 5a Medium IGSCC Medium 1 1 0 no change no change no change no change Si 6a Medium None Low 130 0 -130 negligible negligible negligible negligible Si 7a Low None Low 2 0 -2 negligible negligible negligible negligible Si Total 4.84E-10 5.OOE-10 4.84E-11 5.OOE-11 AF 4 High None Low 5 4 -1 7.OOE-11 7.OOE-11 7.OOE-12 7.OOE-12 AF 5a (3) Medium TT, (FAC) Medium (High) 1 1 0 -1.20E-11 no change -1.20E-12 no change AF 5a Medium TT Medium 1 1 0 -1.20E-11 no change -1.20E-12 no change AF 6a Medium None Low 0 0 0 no change no change no change no change AF Total 4.60E-11 7.00E-11 4.60E-12 7.OOE-12 Page 30 of 44

Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.6-1 (Cont'd)

Unit I - Risk Impact Analysis Results System(') Category Consequence Failure Potential Inspections D CDF Impact'3) LERF Impact(3) lan DMs Rank SXI2 l RI-ISI Delta w/ POD w/o POD w/ POD w/o POD SG 2 (1) High TASCS, (FAC) Medium (High) 3 6 3 -1.26E-08 -4.20E-09 -1.26E-09 -4.20E-10 SG 2 (1) High TT, (FAC) Medium (High) 1 4 3 -9.24E-09 -4.20E-09 -9.24E-10 -4.20E-10 SG 2 High TT Medium 1 2 1 -4.20E-09 -1.40E-09 -4.20E-10 -1.40E-10 SG 4 (1) High None (FAC) Low (High) 4 6 2 -1.40E-10 -1.40E-10 -1.40E-11 -1.40E-11 SG 4 High None Low 10 13 3 -2.10E-10 -2.10E-10 -2.10E-11 -2.1OE-11 SG 5a (3) Medium TT, (FAC) Medium (High) 2 3 1 -4.20E-11 -1.00E-11 -4.20E-12 -1.OOE-12 SG 5a Medium TT Medium 1 1 0 -1.20E-11 no change -1.20E-12 no change SG 6a (3) Medium None (FAC) Low (High) 1 0 -1 negligible negligible negligible negligible SG 6a Medium None Low 22 0 -22 negligible negligible negligible negligible SG Total III_-2.64E-08 -1.02E-08 -2.64E-09 -1.02E-09 Grand Total I I -4.90E-08 -9.17E-09 -4.90E-09 -9.17E-10 Notes

1. Systems are described in Table 3.1-1.

2. Only those ASME Section Xl Code inspection locations that received a volumetric examination in addition to a surface examination are included in the count. Inspection locations previously subjected to a surface examination only were not considered in accordance with Section 3.7.1 of EPRI TR-1 12657.

3. Per Section 3.7.1 of EPRI TR-1 12657, the contribution of low risk categories 6 and 7 need not be considered in assessing the change in risk. They are excluded from analysis because they have an insignificant impact on risk. Hence, the word "negligible" is given in these cases in lieu of values for CDF and LERF Impact. For those cases in high, medium or low risk region piping where no impact to CDF or LERF exists, "no change' is listed.

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Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.6-2 Unit 2 - Risk Impact Analysis Results System' ) Category Consequence Rank I Failure Potential VDMs i Rank lSX(2)

Inspections lRl-S Delta CDF Impact(3 w/ POD w/o POD

[ wi LERF POD lmpact(3) w/o POD RC 2 High TASCS, TT Medium 4 2 -2 -1.68E-09 2.80E-09 -1.68E-10 2.80E-10 RC 2 High TT, PWSCC Medium 3 3 0 no change no change no change no change RC 2 High TASCS Medium 5 3 -2 -3.36E-09 2.80E-09 -3.36E-10 2.80E-10 RC 2 High TT Medium 5 2 -3 -8.40E-10 4.20E-09 -8.40E-1 I 4.20E-10 RC 2 High PWSCC Medium 6 6 0 no change no change no change no change RC 4 High None Low 35 29 -6 4.20E-10 4.20E-10 4.20E-11 4.20E-11 RC 7a Low None Low 0 0 0 no change no change no change no change RC Total -5.46E-09 1.02E-08 -5.46E-10 1.02E-09 CH 2 High TASCS, TT Medium 0 2 2 -5.04E-09 -2.80E-09 -5.04E-10 -2.80E-10 CH 2 High TT Medium 0 5 5 -1.26E-08 -7.OOE-09 -1.26E-09 -7.OOE-10 CH 4 High None Low 0 0 0 no change no change no change no change CH 6a Medium None Low 8 0 -8 negligible negligible negligible negligible CH Total _ -1.76E-08 -9.80E-09 -1.76E-09 -9.80E-10 Si 4 High None Low 15 8 -7 4.90E-10 4.90E-10 4.90E-11 4.90E-1I Si 5a Medium TASCS Medium 2 1 -1 -6.OOE-12 1.OOE-11 -6.OOE-13 1.OOE-12 Si 5a Medium IGSCC Medium 0 1 1 -1.OOE-11 -1.00E-11 -1.OOE-12 -1.00E-12 Si 6a Medium None Low 128 0 -128 negligible negligible negligible negligible Si 7a Low None Low 2 0 -2 negligible negligible negligible negligible Si Total 4.74E-10 4.90E-10 4.74E-11 4.90E-11 AF 4 High None Low 6 4 -2 1.40E-10 1.40E-10 1.40E-11 1.40E-11 AF 5a (3) Medium TT, (FAC) Medium (High) 1 1 0 -1.20E-11 no change -1.20E-12 no change AF 5a Medium TT Medium 1 1 0 -1.20E-11 no change -1.20E-12 no change AF 6a Medium None Low 0 0 0 no change no change no change AF Total I_1.16E-10 1.40E-10 1.16E-11 1.40E-11 Page 32 of 44

Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.6-2 (Cont'd)

Unit 2 - Risk Impact Analysis Results it) 1 C Consequence 1 Failure Potential l Inspections CDF Impact(3) J LERF lmpact(3)

System Category Rank [ VDMs l Rank SXI(2 l RI-ISI l Delta w/ POD w/o POD l w POD w/o POD SG 2 (1) High TASCS, TT, (FAC) Medium (High) 0 1 1 -2.52E-09 -1.40E-09 -2.52E-10 -1.40E-10 SG 2 (1) High TASCS, (FAC) Medium (High) 1 1 0 -1.68E-09 no change -1.68E-10 no change SG 2 (1) High TT, (FAC) Medium (High) 1 2 1 -4.20E-09 -1.40E-09 -4.20E-10 -1.40E-10 SG 2 High TASCS, TT Medium 5 4 -1 -5.88E-09 1.40E-09 -5.88E-10 1.40E-10 SG 2 High TT Medium 1 4 3 -9.24E-09 -4.20E-09 -9.24E-10 -4.20E-10 SG 4 (1) High None (FAC) Low (High) 4 5 1 -7.OOE-1 1 -7.OOE-1 1 -7.OOE-12 -7.OOE-12 SG 4 High None Low 14 18 4 -2.80E-10 -2.80E-10 -2.80E-11 -2.80E-11 SG 5a (3) Medium TT, (FAC) Medium (High) 3 4 1 -5.40E-11 -1.OOE-11 -5.40E-12 -1.OOE-12 SG 5a Medium TT Medium 1 1 0 -1.20E-11 no change -1.20E-12 no change SG 6a (3) Medium None (FAC) Low (High) 1 0 -1 negligible negligible negligible negligible SG 6a Medium None Low 26 0 -26 negligible negligible negligible negligible SG Total _ _ -2.39E-08 -5.96E-09 -2.39E-09 -5.96E-10 Grand Total -4.64E-08 *4.91 E-09 -4.64E-09 .4.91 E-1O0 Notes

1. Systems are described in Table 3.1-2.

2. Only those ASME Section Xl Code inspection locations that received a volumetric examination in addition to a surface examination are included in the count. Inspection locations previously subjected to a surface examination only were not considered in accordance with Section 3.7.1 of EPRI TR-1 12657.

3. Per Section 3.7.1 of EPRI TR-112657, the contribution of low risk categories 6 and 7 need not be considered in assessing the change in risk. They are excluded from analysis because they have an insignificant impact on risk. Hence, the word 'negligible' is given in these cases in lieu of values for CDF and LERF Impact. For those cases in high, medium or low risk region piping where no impact to CDF or LERF exists, 'no change" is listed.

Page 33 of 44

Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.6-3 Unit 3 - Risk Impact Analysis Results System(l) Category Consequence Failure Potential Inspections CDF Impact(3) LERF Impact(3)

Rank DMs Rank SXI(2) R-ISI Delta w/ POD w/o POD w/ POD w/o POD RC 2 High TASCS, TT Medium 4 2 -2 -1.68E-09 2.80E-09 -1.68E-10 2.80E-10 RC 2 High TT, PWSCC Medium 3 3 0 no change no change no change no change RC 2 High TASCS Medium 5 3 -2 -3.36E-09 2.80E-09 -3.36E-10 2.80E-10 RC 2 High TT Medium 5 2 -3 -8.40E-10 4.20E-09 -8.40E-11 4.20E-10 RC 2 High PWSCC Medium 6 6 0 no change no change no change no change RC 4 High None Low 35 29 -6 4.20E-10 4.20E-10 4.20E-1 1 4.20E-1 1 RC 7a Low None Low 0 0 0 no change no change no change no change RC Total -5.46E-09 1.02E-08 -5.46E-10 1.02E-09 CH 2 High TASCS, TT Medium 0 2 2 -5.04E-09 -2.80E-09 -5.04E-10 -2.80E-10 CH 2 High TT Medium 0 5 5 -1.26E-08 -7.00E-09 -1.26E-09 -7.00E-10 CH 4 High None Low 0 0 0 no change no change no change no change CH 6a Medium None Low 8 0 -8 negligible negligible negligible negligible CH Total l ______-1.76E-08 -9.80E-09 -1.76E-09 -9.80E-10 Si 4 High None Low 14 8 -6 4.20E-10 4.20E-10 4.20E-11 4.20E-11 Si 5a Medium TASCS Medium 2 1 -1 -6.00E-12 1.OOE-11 -6.OOE-13 1.OOE-12 Si 5a Medium IGSCC Medium 2 1 -1 1.OOE-11 1.OOE-11 1.OOE-12 1.00E-12 Si 6a Medium None Low 131 0 -131 negligible negligible negligible negligible Si 7a Low None Low 2 0 -2 negligible negligible negligible negligible Si Total _ _ _ _ 4.24E-10 4.40E-10 4.24E-11 4.40E-11 AF 4 High None Low 6 4 -2 1.40E-10 1.40E-10 1.40E-11 1.40E-11 AF 5a (3) Medium TT, (FAC) Medium (High) 2 1 -1 -6.OOE-12 1.00E-1 1 -6.OOE-13 1.00E-12 AF 5a Medium TT Medium 0 1 1 -1.80E-11 -1.OOE-11 -1.80E-12 -1.OOE-12 AF 6a Medium None Low 0 0 0 no change no change no change no change AF Total 1.16E-10 1.40E-10 1.16E-11 1.40E-11 Page 34 of 44

Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 3.6-3 (Cont'd)

Unit 3 - Risk Impact Analysis Results (1) Cto Consequence Failure Potential l Inspections l CDF Impact( 3 ) LERF Impact(3 )

System Category Rank DMs Rank SXI(2) R4-ISI J Delta w/ POD wlo POD w/ POD w/o POD SG 2 (1) High TASCS, (FAC) Medium (High) 3 5 2 -1.01E-08 -2.80E-09 -1.01E-09 -2.80E-10 SG 2 (1) High TT, (FAC) Medium (High) 1 4 3 -9.24E-09 -4.20E-09 -9.24E-10 -4.20E-10 SG 2 High TT Medium 1 2 1 -4.20E-09 -1.40E-09 -4.20E-10 -1.40E-10 SG 4 (1) High None (FAC) Low (High) 4 5 1 -7.OOE-11 -7.OOE-11 -7.OOE-12 -7.OOE-12 SG 4 High None Low 9 12 3 -2.10E-10 -2.1OE-10 -2.10E-11 -2.1 OE-11 SG 5a (3) Medium TT, (FAC) Medium (High) 3 4 1 -5.40E-11 -1.00E-11 -5.40E-12 -1.00E-12 SG 5a Medium TT Medium 1 2 1 -3.OOE-11 -1.OOE-11 -3.OOE-12 -1.OOE-12 SG 6a (3) Medium None (FAC) Low (High) 2 0 -2 negligible negligible negligible negligible SG 6a Medium None Low 21 0 -21 negligible negligible negligible negligible SG Total -2.39E-08 -8.70E-09 -2.39E-09 -8.70E-10 Grand Total -4.64E-08 -7.70E-09 -4.64E-09 -7.70E-10 Notes

1. Systems are described in Table 3.1-3.

2. Only those ASME Section Xl Code inspection locations that received a volumetric examination in addition to a surface examination are included in the count. Inspection locations previously subjected to a surface examination only were not considered in accordance with Section 3.7.1 of EPRI TR-1 12657.

3. Per Section 3.7.1 of EPRI TR-112657, the contribution of low risk categories 6 and 7 need not be considered in assessing the change in risk. They are excluded from analysis because they have an insignificant impact on risk. Hence, the word 'negligible' is given in these cases in lieu of values for CDF and LERF Impact. For those cases in high, medium or low risk region piping where no impact to CDF or LERF exists, 'no change" is listed.

Page 35 of 44

I Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 5-1-1 Unit 1 - Inspection Location Selection Comparison Between ASME Section Xi Code and EPRI TR-112657 by Risk Region 1 High Risk Region Medium Risk Region Low Risk Region 1

System~ ) Code Category Weld Section Xl j____ ____

EPRI TR-112657 Weld Section Xl EPRI TR-112657 Weld Section Xi EPRI TR-112657 Count Vol/Sur Sur Only RI-ISI Other(2) Count Vol/Sur Sur Only RI-ISI Other(2) Count Vol/Sur SurOnly RI-ISI Other(2)

B-F 6 6 0 6 _ _ __ _ _ _ _ _ _ _ _ _ _ _

RC B-JDMWs 3 3 0 3 6 0 6 0 B-J 47 14 5 7 276 35 51 29 8 0 0 0 B-JDMWS 1 0 1 0 CH B-J 23 0 8 7 5 0 1 0 50 0 13 0 C-F-1 85 8 0 0 B-JDMs 4 4 0 4 Si B1- 51 10 9 5 116 16 7 0 C-F-1 43 4 0 1 1434 116 10 0 AF C-F-1 48 7 0 6 11 0 0 0 SG C-F-1 1 0 0 0 26 0 0 0 C-F-2 45 5 0 12 205 17 0 23 294 23 0 0 B-F 6 6 0 6 _

B_JDS 4 3 1 3 10 4 6 4 Total B1- 70 14 13 14 332 45 61 34 174 16 20 0 C-F-1 92 11 0 7 1556 124 10 0 C-F-2 45 5 0 12 205 17 0 23 294 23 0 0 Notes

1. Systems are described in Table 3.1-1.

2. The column labeled 'Other' is generally used to identify plant augmented inspection program locations credited per Section 3.6.5 of EPRI TR-1 12657. The EPRI methodology allows plant augmented inspection program locations to be credited if the inspection locations selected strictly for Rl-ISI purposes produce less than a 10% sampling of the overall Class 1 weld population. This option was not applicable for the PVNGS Rl-ISI application. The 'Other' column has been retained in this table solely for uniformity purposes with the other Rl-ISI application template submittals.

Page 36 of 44

Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 5-1-2 Unit 2 - Inspection Location Selection Comparison Between ASME Section Xl Code and EPRI TR-112657 by Risk Region High Risk Region Medium Risk Region 1 Low Risk Region Category Weld Section XI EPRI TR-112657 Weld Section XI EPRI TR-112657 Weld Section XI EPRI TR-112657 Count Vol/Sur Sur Onlyl RI-ISI Other(2) Count Vol/Sur Sur Only R14SI Other(2) Count Vol/Sur Sur Onl RI-ISI Other(2)

B-F 6 6 0 6 RC B-J DMWs 3 3 0 3 6 0 6 0 B-J 49 14 5 7 264 35 51 29 10 0 1 0 B_jDMWS 1 0 1 _

CH B-J 23 0 8 7 5 0 1 0 47 0 13 0 C-F-1 83 8 0 0 B JDMws 4 4 0 4 Si B-J 46 9 9 5 109 16 7 0 C-F-1 46 4 0 1 1418 114 10 0 AF C-F-1 51 8 0 6 11 0 0 0 SG C-F-1 24 4 0 5 15 4 0 0 C-F-2 20 4 0 7 264 22 0 28 289 23 0 0 B-F 6 6 0 6 B_jDMWS 4 3 1 3 10 4 6 4 .

Total B-J 72 14 13 14 315 44 61 34 166 16 21 0 C-F-1 24 4 0 5 97 12 0 7 1527 126 10 0 G

C-F-2 20 4 0 7 264 22 0 28 289 23 0 0 Notes

1. Systems are described in Table 3.1-2.

2. The column labeled 'Other' is generally used to identify plant augmented inspection program locations credited per Section 3.6.5 of EPRI TR-1 12657. The EPRI methodology allows plant augmented inspection program locations to be credited if the inspection locations selected strictly for Rl-ISI purposes produce less than a 10% sampling of the overall Class 1 weld population. This option was not applicable for the PVNGS Ri-ISI application. The 'Other" column has been retained in this table solely for uniformity purposes with the other RI-ISI application template submittals.

Page 37 of 44

Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 5-1-3 Unit 3 - Inspection Location Selection Comparison Between ASME Section Xl Code and EPRI TR-112657 by Risk Region High Risk Region Medium Risk Region Low Risk Region System() Category Weld Section XI EPRI TR-112657 Weld Section XI EPRI TR-112657 Weld Section XI EPRI TR-112657 Count Vol/Sur Sur Only R14S Oher") Count Vol/Sur Sur Only RI-ISI OtherS2) Count jVol/Sur Sur Only RI-ISI Other(2 )

B-F 6 6 0 6 RC B jDMWS 3 3 0 3 6 0 6 0 B-J 49 14 5 7 265 35 51 29 10 0 1 0 B-JDMWs 1 0 1 0 =

CH B-J 23 0 8 7 5 0 1 0 47 0 13 0 C-F-1 86 8 0 0 B-JDMWs 4 4 0 4 Si B-J 44 10 9 5 107 16 7 0 C-F-1 44 4 0 1 1433 117 10 0 AF C-F-1 48 8 0 6 10 0 0 0 SG C-F-1 = 6 0 0 0 C-F-2 44 5 0 11 209 17 0 23 292 23 0 0 B-F 6 6 0 6 =

B JDMWS 4 3 1 3 10 4 6 4 Total B-J 72 14 13 14 314 45 61 34 164 16 21 0 C-F- 1 92 12 0 7 1535 125 10 0 C-F-2 44 5 0 11 209 17 0 23 292 23 0 0 Notes

1. Systems are described in Table 3.1-3.

2. The column labeled 'Other' is generally used to identify plant augmented inspection program locations credited per Section 3.6.5 of EPRI TR-1 12657. The EPRI methodology allows plant augmented inspection program locations to be credited if the inspection locations selected strictly for RI-ISI purposes produce less than a 10% sampling of the overall Class 1 weld population. This option was not applicable for the PVNGS RI-ISI application. The 'Other' column has been retained in this table solely for uniformity purposes with the other RI-ISI application template submittals.

Page 38 of 44

Risk-Informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 5-2-1 Unit I - Inspection Location Selection Comparison Between ASME Section Xi Code and EPRI TR-112657 by Risk Category l Risk Consequence Failure Potential Code Weld Section Xl EPRI TR.112657 Category Rank Rank DMs Rank Category Count Vol/Sur Sur Onl RI-151 Other( 2 )

RC 2 High High TASCS, TT Medium B-J 15 4 0 2 B-F 2 2 0 2 RC 2 High High TT, PWSCC Medium B13JDM s 1 1 0 1 RC 2 High High TASCS Medium B-J 25 5 4 3 RC 2 High High TT Medium B-J 7 5 1 2 RC 2 High High PWSCC Medium B-F 2 2 0 2 B-jDMWS 6 0 6 0 RC 4 Medium High None Low B-j 6 3 6 2

_ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ -J 13__ 276 35 51 29 RC 7a Low Low None Low B-J 8 0 0 0 CH 2 High High TASCS, TT Medium B-J 2 0 0 2 BJDMWs 1 O 1 O CH 2 High High TT Medium B-J 1 0 1 0

_ _ _ _ _ _ _ __ _ ___ _ _ _ _ _ _B -J 21 0 8 5 CH 4 Medium High None Low B-J 5 0 1 0 B-J 50 0 13 0 CH 6a Low Medium None Low

_ __ __ _ __ __ _ _ __ _ _ _ _ __ _ __ __ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ C-F -I 85 _ 8 0 0 B-JDMWS 4 4 0 4 Si 4 Medium High None Low B-J 39 9 9 4 C-F-1 37 2 0 0 Si5 a Medium Medium TASCS Medium C-F-1 6 2 0 1 Si 5a Medium Medium IGSCC Medium B-J 12 1 0 1 Si 6a Low Medium None Low B-. 116 16 7 0

_ _ C-F-1 1428 114 10 0 Si 7a Low Low None Low C-F-1 6 2 0 0 Page 39 of 44

Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 5-2-1 (Cont'd)

Unit I - Inspection Location Selection Comparison Between ASME Section Xi Code and EPRI TR-112657 by Risk Category Risk Consequence Failure Potential Code Weld Section Xi IEPRI TR-112657 System(') Rank Category Count Category Rank DMs Rank Vol/Sur ISur Onlyj RI-ISI Other(2)

=

AF 4 Medium High None Low C-F-1 38 5 4 AF 5a (3) Medium (High) Medium TT, (FAC) Medium (High) C-F-1 1 1 I AF 5a Medium Medium TT Medium C-F-1 9 1 1 AF 6a Low Medium None Low C-F-1 II 0 0 SG 2 (1) High (High) High TASCS, (FAC) Medium (High) C-F-2 33 3 6 SG 2 (1) High (High) High TT, (FAC) Medium (High) C-F-2 6 1 4 SG 2 High High TT Medium C-F-2 6 1 2 SG 4 (1) Medium (High) High None (FAC) Low (High) C-F-2 52 4 6 SG 4 Medium High None Low C-F-2 122 10 0 13 SG 5a (3) Medium (High) Medium TT, (FAC) Medium (High) C-F-2 23 2 0 0 SG 5a Medium Medium TT Medium C-F-2 8 1 0 1 SG 6a (3) Low (High) Medium None (FAC) Low (High) C-F-2 10 0 0 0 C-F-I 22 0 0 0 SG 6a Low Medium None Low C-F-2 284 22 0 0 Notes

1. Systems are described in Table 3.1-1.

2. The column labeled 'Other' is generally used to identify plant augmented inspection program locations credited per Section 3.6.5 of EPRI TR-1 12657. The EPRI methodology allows plant augmented inspection program locations to be credited if the inspection locations selected strictly for RI-ISI purposes produce less than a 10% sampling of the overall Class 1 weld population. This option was not applicable for the PVNGS Rl-ISI application. The 'Other" column has been retained in this table solely for uniformity purposes with the other Rl-ISI application template submittals.

Page 40 of 44

j Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 5-2-2 Unit 2 - Inspection Location Selection Comparison Between ASME Section Xi Code and EPRI TR-112657 by Risk Category System(1 ) Risk Consequence Failure Potential l Code Weld Section Xl EPRI TR-112657 l Category J

Rank Rank DMs Rank Category Count Vol/Sur Sur Only R1-ISI Other(2)

- =

RC 2 High High TASCS, TT Medium B-J 15 4 0 2 RC 2 High High TT, PWSCC Medium B-.FJDS 2 2 21 RC 2 High High TASCS Medium B-J 27 5 4 3 RC 2 High High TT Medium B-J 7 5 1 2 B-F 4 4 0 4 ___

RC 2 High High PWSCC Medium B-J DM s 2 2 0 2 B-JDM~s 6 0 6 0 ___

RC 4 Medium High None Low B-J 6 3 6 2

________ __ __ __ ___ __ _ _ __ ___ __ __ B -J 264 35 51 29 RC 7a Low Low None Low B-J 10 0 1 0 CH 2 High High TASCS, TT Medium B-J 2 0 0 2 1-j DM~s I 0 1 0 ___

CH 2 High High TT Medium B-J 21 0 8 5 CH 4 Medium High None Low B-J 5 0 1 0 B-J 47 0 13 0 ___

CH 6a Low Medium None Low .

C-F-1 83 8 0 0 B-JDMs 4 4 0 4 Si 4 Medium High None Low B-J 34 9 9 4 C-F-1 40 2 0 0 Si 5a Medium Medium TASCS Medium C-F-I 6 2 0 1 Si 5a Medium Medium IGSCC Medium BJ 12 0 0 1 Si 6a Low Medium None Low B-J 109 16 7 0 C-F-1 1412 112 10 0 Si 7a Low Low None Low C-F-1 6 2 0 0 Page 41 of 44

Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 5-2-2 (Cont'd)

Unit 2 - Inspection Location Selection Comparison Between ASME Section Xi Code and EPRI TR-112657 by Risk Category System(1) Risk Consequence Failure Potential Code Weld Section Xl EPRI TR-112657 Category Rank Rank DMs J Rank Category Count Vol/Sur Sur Only RI-ISI Other(2)

AF 4 Medium High None Low C-F-1 40 6 0 4 AF 5a (3) Medium (High) Medium TT, (FAC) Medium (High) C-F-1 1 1 0 1 AF 5a Medium Medium TT Medium C-F-1 10 1 0 1 AF 6a Low Medium None Low C-F-1 II 0 0 0 SG 2 (1) High (High) High TASCS, TT, (FAC) Medium (High) C-F-1 4 0 0 1 SG 2 (1) High (High) High TASCS, (FAC) Medium (High) C-F-2 4 1 0 1 SG 2 (1) High (High) High TT, (FAC) Medium (High) C-F-2 6 1 0 2 SG 2 High High TASCS, TT Medium C-F-I 20 4 0 4

___ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ C-F-2 4 1 0 0 SG 2 High High TT Medium C-F-2 6 1 0 4 SG 4 (1) Medium (High) High None (FAC) Low (High) C-F-2 48 4 0 5 SG 4 Medium High None Low C-F-2 174 14 0 18 SG 5a (3) Medium (High) Medium TT, (FAC) Medium (High) C-F-2 32 3 0 4 SG 5a Medium Medium TT Medium C-F-2 10 1 0 1 SG 6a (3) Low (High) Medium None (FAC) Low (High) C-F-2 3 1 0 0 C-F-I 15 4 0 0 SG 6a Low Medium None Low

. C-F-2 286 22 0 0 Notes

1. Systems are described in Table 3.1-2.

2. The column labeled 'Other' is generally used to identify plant augmented inspection program locations credited per Section 3.6.5 of EPRI TR-112657. The EPRI methodology allows plant augmented inspection program locations to be credited if the inspection locations selected strictly for RI-ISI purposes produce less than a 10% sampling of the overall Class 1 weld population. This option was not applicable for the PVNGS RI-ISI application. The 'Other' column has been retained in this table solely for uniformity purposes with the other RI-ISI application template submittals.

Page 42 of 44

Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 5-2-3 Unit 3 - Inspection Location Selection Comparison Between ASME Section Xl Code and EPRI TR-112657 by Risk Category System( [ Risk Consequence l Failure Potential Code Weld Section XI lEPRI TR-112657 System Category l Rank Rank DMs Rank Category Count Vol/Sur Sur Onll Rl-ISI lOther2)

RC 2 High High TASCS, TT Medium B-J 15 4 0 2 B-F 2 2 0 2 ___

RC 2 High High TT, PWSCC Medium BDFs 1 2 0 1 RC 2 High High TASCS Medium B-J 27 5 4 3 RC 2 High High TT Medium B-J 7 5 1 2 RC 2 High High PWSCC Medium BFJDws 2 2 0 2 RC 4 Medium High None Low gJDMs 6 0 6 0 3B-J 265 35 51 29 RC 7a Low Low None Low B-J 10 0 1 0 CH 2 High High TASCS, TT Medium B-J 2 0 0 2 CH 2 High High TT Medium B-JDWS I 0 1 05

__ _ _ _ _ _ _ _ _ _ __ __ __ __ __ _ _ _ _ _ _ _ _ _ _3-J 21 0 8 5 _ _ _

CH 4 Medium High None Low B-J 5 0 1 0 CH 6a Low Medium None Low B-J 47 O 13 O

_C-_-F-1 86 8 0 0 1-J DM~s 4 4 0 4 Si 4 Medium High None Low 1B-J 32 8 9 4

.C-F- 38 2 0 0 Si 5a Medium Medium TASCS Medium C-F-1 6 2 0 1 Si 5a Medium Medium IGSCC Medium B-J 12 2 0 1 Si 6a Low Medium None Low B-J 107 16 7 O C-F-1 1427 115 10 0 Si 7a Low Low None Low C-F-1 6 2 0 0 Page 43 of 44

Risk-informed Inservice Inspection Program Plan, Palo Verde Nuclear Generating Station, Revision 0 Table 5-2-3 (Cont'd)

Unit 3 - Inspection Location Selection Comparison Between ASME Section Xl Code and EPRI TR-112657 by Risk Category Sytem 11 Category Risk l Rank 1 Consequence Rank l

DMs MFailure Potential l Rank l Code Category l Weld Count Section XI Vol/Sur lSur Only EPRI TR-112657 RI-ISI lOther(2)

System AF 4 Medium High None Low C-F-1 38 6 0 4 AF 5a (3) Medium (High) Medium TT, (FAC) Medium (High) C-F-1 2 2 0 1 AF 5a Medium Medium TT Medium C-F-1 8 0 0 1 AF 6a Low Medium None Low C-F-1 10 0 0 0 SG 2 (1) High (High) High TASCS, (FAC) Medium (High) C-F-2 32 3 0 5 SG 2 (1) High (High) High TT, (FAC) Medium (High) C-F-2 4 1 0 4 SG 2 High High TT Medium C-F-2 8 1 0 2 SG 4 (1) Medium (High) High None (FAC) Low (High) C-F-2 46 4 0 5 SG 4 Medium High None Low C-F-2 114 9 0 12 SG 5a (3) Medium (High) Medium TT, (FAC) Medium (High) G-F-2 37 3 0 4 SG 5a Medium Medium TT Medium C-F-2 12 1 0 2 SG 6a (3) Low (High) Medium None (FAC) Low (High) C-F-2 23 2 0 0 SG 6a Low Medium None Low G-F-2 269 21 0 0 Notes

1. Systems are described in Table 3.1-3.

2. The column labeled 'Other' is generally used to identify plant augmented inspection program locations credited per Section 3.6.5 of EPRI TR- 12657. The EPRI methodology allows plant augmented inspection program locations to be credited if the inspection locations selected strictly for RI-ISI purposes produce less than a 10% sampling of the overall Class 1 weld population. This option was not applicable for the PVNGS RI-ISI application. The 'Other' column has been retained in this table solely for uniformity purposes with the other RI-ISI application template submittals.

Page 44 of 44

ML060310511

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