HNP-05-049, Relief Request to Use a Risk-Informed Inservice Inspection Program for Class 1 and 2 Piping Welds

Relief Request to Use a Risk-Informed Inservice Inspection Program for Class 1 and 2 Piping Welds

ML051230314
Person / Time
Site:	Harris
Issue date:	04/27/2005
From:	Morton T Progress Energy Carolinas, Progress Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
HNP-05-049
Download: ML051230314 (29)

Text

_Ir Progress Energy Serial: HNP-05-049 APR 2 7 2005 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 RELIEF REQUEST TO USE A RISK-INFORMED INSERVICE INSPECTION PROGRAM FOR CLASS 1 AND 2 PIPING WELDS Ladies and Gentlemen:

In accordance with the Code of Federal Regulations, Title 10, Part 50.55a, "Codes and Standards," paragraph (a)(3)(i), the Harris Nuclear Plant (HNP) of Carolina Power and Light Company (CP&L) doing business as Progress Energy Carolinas, Inc., requests relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components," requirements for the selection and examination of Class I and 2 piping welds. HNP proposes to use the alternative methodology contained in the NRC-approved Electric Power Research Institute (EPRI) Topical Report (TR) 112657, Revision B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure."

HNP requests approval of this relief request, pursuant to 10 CFR 50.55a(3)(i), since the proposed alternative would provide an acceptable level of quality and safety. provides the proposed HNP relief request (2RG-01 0). provides the proposed HNP Risk-informed Inservice Inspection Program Plan.

HNP requests approval of this relief request by November 11, 2005 to support planning activities associated with Refueling Outage (RFO) 13 scheduled for the spring of 2006.

Please refer any questions regarding this submittal to Mr. Dave Corlett at (919) 362-3137.

Sincerely, T. C. Morton Manager, Site Support Services Progress Energy Carolinas, Inc.

Harris Nuclear Plant P.O. Box 165 New Hill, NC 27562

HNP Serial No. HNP-05-049 Page 2 TCM/jpy C: Mr. R. A. Musser, NRC Sr. Resident Inspector Mr. C. P. Patel, NRC Project Manager Dr. W. D. Travers, NRC Regional Administrator

to :"I ATTACHMENT 1-TO SERIAL: HNP-05-049 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. I NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 INSERVICE INSPECTION PROGRAM RELIEF REQUEST NO. 2RG-010 REQUEST TO USE RISK-INFORMED INSERVICE INSPECTION FOR CLASS 1 AND 2 PRESSURE-RETAINING PIPING WELDS COMPONENTS FOR WHICH RELIEF IS REQUESTED:

Code Class: Class 1 and 2

References:

ASME, Section Xl, 1989 Edition with no Addenda, Subarticles IWB-2500 and IWC-2500, and Tables IWB-2500-1 and IWC-2500-1 Examination Categories: B-F, B-J, C-F-1, and C-F-2

==

Description:==

All Class 1 and 2 pressure-retaining piping welds CODE REQUIREMENTS:

American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, 1989 Edition with no Addenda, Subarticle IWB-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, Categories B-F and B-J require 100% and 25% respectively of the total number of non-exempt welds.

Subarticle IWC-2500 (a) states:

"Components shall be examined and pressure 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, Categories C-F-1 and C-F-2 require 7.5%, but not less than 28 welds to be selected for examination.

In addition, both Tables (IWB-2500-1 and IWC-2500-1) reference figures that convey the examination volume for each configuration that could be encountered.

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ATTACHMENT I TO SERIAL: HNP-05-049 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 INSERVICE INSPECTION PROGRAM RELIEF REQUEST NO. 2RG-010 REQUEST TO USE RISK-INFORMED INSERVICE INSPECTION FOR CLASS 1 AND 2 PRESSURE-RETAINING PIPING WELDS BASIS FOR RELIEF:

Harris Nuclear Plant (HNP) requests relief in accordance with 10 CFR 50.55a(a)(3)(i) since the proposed alternative would provide an acceptable level of quality and safety.

The Nuclear Regulatory Commission has approved previously several Risk-Informed Inservice Inspection (RI-ISI) Programs based on the methodology contained in Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Revision B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure." Similar RI-ISI submittals have been approved recently for Cooper and Salem (Units 1 and 2) Nuclear Plants.'

ALTERNATE EXAMINATIONS:

As an alternative to the existing ASME Section Xl requirements for selection and examination of Class I and 2 piping welds, HNP will implement the alternative methods of a RI-ISI program as specified by EPRI TR-112657, Revision B-A and detailed in of this letter.

TECHNICAL JUSTIFICATION FOR REQUESTING RELIEF:

The scope for ASME Section Xl Inservice Inspection (ISI) programs is largely based on deterministic results contained in design stress reports. These reports are normally very conservative and may not be an accurate representation of failure potential. As allowed by 10 CFR 50.55a(b)(2)(ii), HNP has been using the alternative selection methodology of the 1974 Edition through Surnmer 1975 Addenda of ASME Section XI for Category B-J (Pressure Retaining Welds in Piping) welds. Service experience has shown that failures are due to either corrosion or fatigue and typically occur in areas not included in the plant's ISI program. Consequently, nuclear plants are devoting significant resources to inspection programs that provide minimum benefit.

As an alternative, significant industry attention has been devoted to the application of risk-informed selection criteria to determine the scope of ISI programs at nuclear power plants. EPRI studies indicate that the application of these techniques will allow operating nuclear plants to reduce the examination scope of current ISI programs by as much as 60% to 80%, significantly reduce costs and radiation exposure, and continue to maintain an equivalent or better safety level as current ASME Section Xl selection criteria.

1. Letter from M. Webb (NRC) to R. Edington (NPPD), dated December 9, 2004, TAC No. MC2351; and Letter from J. Clifford (NRC) to R. Anderson (PSEG Nuclear),

dated October 1, 2003, TAC NOS. MB7537 and MB7538, respectively.

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ATTACHMENT 1 TO SERIAL: HNP-05-049 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 INSERVICE INSPECTION PROGRAM RELIEF REQUEST NO. 2RG-010 REQUEST TO USE RISK-INFORMED INSERVICE INSPECTION FOR CLASS 1 AND 2 PRESSURE-RETAINING PIPING WELDS TECHNICAL JUSTIFICATION FOR REQUESTING RELIEF (continued):

HNP has applied the methodology of EPRI TR-112657 Revision B-A in the development of the proposed HNP RI-ISI Program (see Attachment 2 of this letter). The use of this methodology for the selection and subsequent examination of Class 1 and 2 piping welds will provide an acceptable level of quality and safety.

IMPLEMENTATION SCHEDULE:

Approval of this alternative is requested for the remainder of the HNP Second 10-year ISI interval.

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ATTACHMENT 2 TO SERIAL: HNP-05-049 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 INSERVICE INSPECTION PROGRAM RELIEF REQUEST NO. 2RG-010 REQUEST TO USE RISK-INFORMED INSERVICE INSPECTION FOR CLASS 1 AND 2 PRESSURE-RETAINING PIPING WELDS Risk-informed Inservice Inspection Program Plan Harris Nuclear Plant, Revision A Page A2-1 of 24

RISK-INFORMED INSERVICE INSPECTION PROGRAM PLAN HARRIS NUCLEAR PLANT, REVISION A Table of Contents

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

2. Proposed Alternative to Current Inservice Inspection Programs 2.1 ASME Section Xl 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 ISI Program Plan Change

6. References/Documentation Page A2-2 of 24

1. INTRODUCTION The Harris Nuclear Plant (HNP) is currently in the second inservice inspection (ISI) interval as defined by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Section Xl Code for Inspection Program B. HNP plans to start implementing a risk-informed inservice inspection (RI-ISI) program during the third inspection period. The ASME Section Xl Code of Record for the second ISI interval at HNP is the 1989 Edition.

The objective of this submittal is to request the use of a risk-informed process for the inservice inspection of Class 1 and 2 piping. The RI-ISI process used in this submittal is described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A 'Revised Risk-Informed Inservice Inspection Evaluation Procedure." 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."

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 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis" and Regulatory Guide 1.178, 'An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping". Further information is provided in Section 3.6.2 relative to defense-in-depth.

1.2 PSA Quality The HNP PSA model inputs used for this application were generated using updated Individual Plant Examination (IPE) models developed in response to Generic Letter (GL) 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities," and associated supplements, submitted to the NRC by letter dated August 20, 1993. The NRC staff SER for the IPE concluded that the study fully met the intent of the Generic Letter 88-20 with no areas of improvement identified. However, a procedure change was implemented to provide for manual operation of circuit breakers if the non-vital 125 VDC control power is lost. The current HNP PSA model, based on the IPE, is an internal events model that includes contribution from internal flooding. The PSA models both Level 1 (CDF) and Level 2 (LERF) utilizing a fault tree linking approach with small functional event trees and detailed, linked fault trees.

The HNP PSA model is detailed, including a variety of initiating events, modeled systems, operator actions, and common cause events. The HNP PSA model reflects the as-built and as-operated plant and explicitly models a number of internal initiating events such as: general transients, LOCAs, support system failures, and internal flooding events. The HNP PSA model explicitly models a number of frontline and support systems, typical of other Westinghouse plants, which are credited in the accident sequence analyses. A number of support system initiating events are modeled with fault trees and linked directly into the HNP PSA model. The HNP PSA model explicitly models a large number of common cause component failures.

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The HNP PSA model has been updated to reflect changes in plant configuration and operator response to events and to address changes in key plant component performance. The HNP PSA model is periodically updated in accordance with requirements set forth in PGN procedure ADM-NGGC-0004 to incorporate any PSA significant plant modifications, to update the various plant initiating event frequencies and to incorporate component operating history for major components into the failure rate database. Computer programs that process PSA model inputs are verified and validated in accordance with administrative procedure CSP-NGGC-2505, "Software Quality Assurance and Configuration Control of Business Computer Systems." This procedure provides for software verification and validation to ensure the software meets the software requirement specifications and functional requirements.

The Harris IPE model of 1993 was updated in 1995, 1997, 1998, 2000, 2001, and 2003 to reflect the changes made to the plant since the development of the original submittal.

A review of these updates has identified the following significant changes to the model:

• Credited local operation of TDAFW pump when DC power is unavailable

• Plant specific data for major pumps and diesels updated through 2002

• Initiating events plant-specific data updated through 2002

• Added system fault trees for Demineralized Water and Main Feedwater/Condensate

• Incorporated operator recovery actions including the alignment of offsite ac breakers, alignment and restoration of MFW after trip, alignment of alternative fuel oil supplies for the EDGs, alignment of swing standby pumps for CCW and HHSI, and operator actuation of the ESW system on failure of NSW return valves

• Incorporated air compressor system modification (replacement compressors) and developed Instrument Air Initiating Event fault tree

• Incorporated plant modifications and procedure changes due to the replacement of steam generators and power uprate of approximately 4.5%

The HNP PSA model is continually implemented and studied by PSA personnel in the performance of their duties. Electronic copies of the models are maintained in a controlled read-only server location. Potential model modifications/enhancements are captured in the corrective action program, evaluated, and maintained for further investigation and subsequent implementation, if necessary. Each supporting element of the HNP PSA model is documented, typically in a stand-alone report. Each analysis element is reviewed by cognizant personnel and comments reconciled before final approval.

The HNP PSA model received a formal industry PRA Peer Review based on the NEI guidelines in June 2002. The WOG Peer Review of the HNP PSA model was conducted by a diverse group of PSA engineers from other PWR plants and industry. PSA-consultants familiar with the PWR plant design. The certification review covered all aspects of the HNP PSA model and the administrative processes used to maintain and update the model. This review generated specific recommendations for model changes, as well as guidance for improvements to processes and methodologies used in the HNP PSA model, and enhancements to the documentation of the model and the administrative procedures used for model updates.

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The HNP PSA model has been revised to address the majority of the issues identified during the peer review. Internal Flooding, HRA update, MAAP re-analysis of specific success criteria, and system engineer review of PSA notebooks are four areas currently being addressed with plans to incorporate any changes into the model in 2005.

Significant changes made to the model resulting from peer review include:

• ATWS model revised to implement the latest Westinghouse guidance in WCAP-15831

• Truncation analysis study performed and model truncation lowered to 1E-10

• Implemented Rhodes Seal LOCA model

• Incorporated the latest Westinghouse guidance for SGTR modeling (WCAP-1 5955)

Two items that remain outstanding, room heat-up analysis documentation, and ISLOCA consideration of RHR HX Channel Head failure, are planned to be addressed in 2005.

The HNP PSA model used in the RI-ISI consequence evaluation is HNP PSA Model of Record - MOR2003. The base case Core Damage Frequency (CDF) is 2.47E-5/year and the Large Early Release Frequency (LERF) is 4.09E-6/year. These numbers are generally consistent with the industry average.

2. PROPOSED ALTERNATIVE TO CURRENT ISI PROGRAMS 2.1 ASME Section Xl ASME Section Xl 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. The alternative RI-ISI Program for piping is described in EPRI TR-112657.

The 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) by alternatively providing an acceptable level of quality and safety. Other non-related portions of the ASME Section Xl Code will be unaffected. EPRI TR-112657 provides the requirements for defining the relationship between the RI-ISI Program and the remaining unaffected portions of ASME Section XI.

2.2 Augmented Programs The following plant augmented inspection programs were considered during the RI-ISI application:

A plant augmented inspection program is currently implemented in response to SER Question 250.1 that requires volumetric or surface examination of thin-walled charging, containment spray, residual heat removal and safety injection system piping welds. The piping of concern addressed by IE Bulletin 79-17, 'Pipe Cracks in Stagnant Borated Water Systems at PWR Plants", in included in the scope of this plant augmented inspection program. The intergranular stress corrosion cracking concern addressed by IE Bulletin 79-17 was explicitly considered in the application of the EPRI RI-ISI process and is subsumed by the RI-ISI Program. The piping Page A2-5 of 24

addressed by SER Question 250.1 was assessed as part of the RI-ISI application scope for HNP. As a result, the RI-ISI Program effectively subsumes this plant augmented inspection program.

• The plant augmented inspection program for high energy line breaks outside containment, implemented in response to Paragraphs B - F of FSAR 6.6.8, "Augmented Inservice Inspection To Protect Against Postulated Piping Failures", is not affected or changed by the RI-ISI Program.

• A plant augmented inspection program was implemented during the first interval in response to NRC Bulletin 88-08, 'Thermal Stresses in Piping Connected to Reactor Coolant Systems". Augmented volumetric examinations were performed on selected piping welds in the charging and safety injection systems considered potentially subject to thermal fatigue. The specific weld locations inspected during the first interval have been instrumented to monitor thermal variations. The thermal fatigue concern addressed by NRC Bulletin 88-08 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 flow accelerated corrosion (FAC) per Generic Letter 89-08 is relied upon to manage this damage mechanism but is not otherwise affected or changed by the RI-IS I Program.

3. RISK-INFORMED ISI PROCESS The process used to develop the RI-ISI Program conformed to the methodology described in EPRI TR-112657 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 for HNP. Table 3-16 of EPRI TR-112657 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 Page A2-6 of 24

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

> AT > 500F, 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 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 if the 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. If there 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.

> 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.

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> 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 (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"). The methodology used in the HNP 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 HNP RI-ISI application if warranted. It should be noted that the NRC has granted approval for RI-ISI relief requests incorporating these TASCS criteria at several facilities, including Comanche Peak (SER dated September 28, 2001) and South Texas Project (SER dated March 5, 2002).

3.1 Scope of Program The systems included in the RI-ISI Program are provided in Table 3.1. The piping and instrumentation diagrams and additional plant information including the existing plant ISI Program were used to define the Class I 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-112657.

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-112657, with the exception of the previously stated deviation.

Table 3.3 summarizes the failure potential assessment by system for each degradation mechanism that was identified as potentially operative.

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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 Table 3.4.

3.5 Element and NDE Selection In general, EPRI TR-112657 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-112657, 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 HNP, the percentage of Class 1 piping welds selected strictly for RI-ISI purposes was 8.3%. 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 increases to 10.1%.

A brief summary is provided in the following table, and the results of the selections are presented in Table 3.5. Section 4 of EPRI TR-112657 was used as guidance in determining the examination requirements for these locations.

Unit Class I Piping Welds(1 )

U Total l Selected 1 624 1 52 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.

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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 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-112657 will be followed.

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

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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-112657, 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-112657 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.

HNP conducted a risk impact analysis per the requirements of Section 3.7 of EPRI TR-112657. The analysis estimates the net change in 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-112657. 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 (3.5E-02) and CLERP (1E-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 XO and is expected to have a value less than IE-08. Piping locations identified as medium failure potential have a likelihood of 20x0 . These PBF likelihoods are consistent with References 9 and 14 of EPRI TR-112657. In addition, the analysis was performed both with and without taking credit for enhanced inspection effectiveness due to an increased POD from application of the RI-ISI approach.

Table 3.6-1 presents a summary of the RI-ISI Program versus 1989 ASME Section Xl Code Edition program requirements and identifies on a per system basis each applicable risk category. 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 Page A2-11 of 24

. I 1I 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, Table 3.6-1 presents 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 delta risk assessment for the Arkansas Nuclear One, Unit 2 (ANO-2) pilot application. An example is provided below.

System S Category Risk Rank")

[ Consequence Rank DMs Failure Potential Rank 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).

FW ,

. 5 (3) t Medium (Hh)

I Medium Kffi

' TASCS, TT, (FAC) ' Medium (High)

.' . X ,I ,*

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 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 Table 3.6-1 but it is Included in Table 5-2.

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

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Risk Impact Results Systems ) ARiSkcDF ARiALERFl wi POD R wlo POD wi POD Iw/oLPOD RC 1.61E-08 7.91E-08 4.60E-10 2.26E-09 RHR -3.68E-10 -3.60E-10 -1.18E-11 -1.1 OE-11 SI -1.79E-09 -1.79E-09 -5.40E-11 -5.40E-11 Cs -6.67E-09 -3.86E-09 -1.92E-10 -1.11E-10 MS negligible negligible negligible negligible FW -1.20E-11 negligible -1.20E-12 negligible CT 1.75E-10 1.75E-10 5.OOE-12 5.00E-12 AF -1.20E-11 negligible -1.20E-12 negligible Total 7.43E-09 7.33E-08 2.05E-10 2.09E-09 Note

1. Systems are described In Table 3.1.

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, that is, a determination of each location's susceptibility to degradation and secondly, an independent assessment of the consequence of the piping failure. These two ingredients assure defense in depth is maintained. First, by evaluating a location's susceptibility to degradation, the likelihood of finding flaws or indications that may be precursors to leak or ruptures is increased. Secondly, 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.

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4. IMPLEMENTATION AND MONITORING PROGRAM Upon approval of the RI-ISI Program, procedures that comply with the guidelines described in EPRI TR-112657 will be prepared to implement and monitor the program. The new program will be integrated into the second inservice inspection interval. No changes to the Technical Specifications are necessary but some changes to the Final Safety Analysis Report will be required for program implementation.

The applicable aspects of the ASME Code not affected by this change will be retained, such as inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements. Existing ASME Section Xi program implementing procedures will be retained and modified to address the 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 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 adjusted on an ASME period basis. Inaddition, 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 RI-ISI Program and ASME Section Xl 1989 Code Edition program requirements for in-scope piping is provided in Tables 5-1 and 5-2. Table 5-1 provides a summary comparison by risk region. Table 5-2 provides the same comparison information, but in a more detailed manner by risk category, similar to the format used in Table 3.6-1.

HNP plans to start implementing a risk-informed inservice inspection (RI-ISI) program during the third inspection period. Beginning with the third inspection period, inspection locations selected per the RI-ISI process will replace those formerly selected per ASME Section Xl criteria. By the end of the second period, 64.8% of the piping weld examinations required by ASME Section Xl have been completed thus far in the second ISI interval for Examination Categories B-F, B-J, C-F-1 and C-F-2. To ensure the performance of 100% of the required examinations during the current ten-year ISI interval, 35.2% of the inspection locations selected for examination per the RI-ISI process will be examined in the third period.

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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 XI, paragraphs IWB-2412 and IWC-2412 are met.

6. REFERENCESIDOCUMENTATION EPRI TR-112657, "Revised Risk-Informed Inservice Inspection Evaluation Procedure", Rev. B-A ASME Code Case N-578, "Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method B,Section XI, Division 1" Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis" Regulatory Guide 1.178, "An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping" Supporting Onsite Documentation HNP-01Q-302, "Consequence Evaluation for Harris Nuclear Plant", Revision 1 HNP-01 Q-301, "Degradation Mechanism Evaluation for the Class 1 and Class 2 Piping Welds at Harris Nuclear Plant", Revision I AR-133830-01, "Service History Review for Harris Nuclear Plant", Revision 0 HNP-01 Q-303, "Risk Ranking for Harris Nuclear Plant", Revision 0 HNP-01Q-304, "Minutes of the Element Selection Meeting for the RI-ISI Project at Harris Nuclear Plant", Revision 0 HNP-01Q-305, "Risk Impact for Harris Nuclear Plant", Revision 0 Page A2-15 of 24

Table 3.1 System Selection and Segment / Element Definition System Description Number of Segments Number of Elements RC - Reactor Coolant System 73 336 RHR - Residual Heat Removal System 24 231 Si - Safety Injection System 84 843 CS - Charging System 32 241 MS - Main Steam System 29 103 FW - Feedwater System 21 79 CT - Containment Spray System 32 299 AF - Auxiliary Feedwater System 12 133 Totals 307 2265 Page A2-16 of 24

Table 3.3 Failure Potential Assessment Summary System~1 ) Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC RC X X X RHR X Si x x__ _ _ __ _ _

FW X _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _X l FCT X l X lllllll _ X Note

1. Systems are described in Table 3.1.

Page A2-17 of 24

Table 3.4 Number of Segments by Risk Category With and Without Impact of FAC High Risk Region Medium Risk Region Low Risk Region System(1 Category I Category 2 Category 3 Category 4 Category 5 Category 6 l Category 7 With Without With ( Without j With Without With ] Without j With J Without With l Without With J Without RC 28 28 42 42 3 3 RHR 2 2 4 4 18 188 Si 9 9 18 18 36 36 21 21 CS 1 1 4 4 3 3 23 23 1 1 Ms _ _ _ _ _ _ _ _ _ _ _ _ 29 29 _

FW 15(2) 0 3 3 3 18 CT = =_=- ==_4 4 4 4 24 24 AF 29 29 15 __61_61 3 3 9 9 Total 29 29 15 0 61 61 31 31 122 137 49 49 Notes

1. Systems are described in Table 3.1.

2. These fifteen segments become Category 6 after FAC is removed from consideration due to no other damage mechanisms being present.

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Table 3.5 Number of Elements Selected for Inspection by Risk Category Excluding Impact of FAC

_ High Risk Region Medium Risk Region Low Risk Region System1) Category 1 Category 2 Category 3 Category 4 Category 5 Category 6 Category 7 Total Selected Total lSelected Total Selected Total Selected Total lSelect ed Total l Selected Total JSelected RC 57 15 273 28 6 l RHR _ __I____ 16 2 6 1 209° Si 225 23 36 4 387 0 195 0 CS 2 1 29 3 8 1 194 0 8 0 MS 103 0 FW _ 1 73 0 CT 15 2 49 0 235 0 AF 6 1 127 0 Total 59 16 558 58 62 8 1142 0 444 0 Notes

1. Systems are described In Table 3.1.

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Table 3.6-1 Risk Impact Analysis Results System1) Category Consequence Failure Potential Inspections CDF Impact( 3 1 LERF Impact(3)

Rank DMs Rank SXI(2) J RlSI l Delta wl POD wlo POD w/ POD w/o POD RC 2 High TASCS, TT Medium 16 4 -12 8.40E-09 4.20E-08 2.40E-10 1.20E-09 RC 2 High TT, PWSCC Medium 1 1 0 no change no change no change no change RC 2 High TASCS Medium 9 5 -4 -1.26E-08 1.40E-08 -3.60E 4.00E-10 RC 2 High Tr Medium 2 0 -2 4.20E-09 7.OOE-09 1.20E-10 2.00E-10 RC 2 High PWSCC Medium 8 5 -3 1.05E-08 1.05E-08 3.00E-10 3.00E-10 RC 4 High None Low 60 28 -32 5.60E-09 5.60E-09 1.60E-10 1.60E-10 RC 7a Low None Low 0 0 0 no change no change no change no change RC Total 1.61 E-08 7.91 E-08 4.60E-10 2.26E-09 RHR 4 High None Low 0 2 2 -3.50E-10 -3.50E-10 -1.002-11 -1.OOE-11 RHR 5a Medium TASCS Medium 0 1 1 -1.80E-11 -1.00E-11 -1.80E-12 -1.OOE-12 RHR 6a Medium None Low 5 0 -5 negligible negligible negligible negligible RHR Total -3.68E-10 -3.60E-10 -1.18E-11 -1.1012-11 Si 4 High None Low 13 23 10 -1.75E-09 -1.752-09 -5.OOE-11 -5.OOE-11 Si 5a Medium Tr, IGSCC Medium 0 2 2 -2.OOE-11 -2.00E-11 -2.OOE-12 -2.OOE-12 Si 5a Medium TT Medium 0 0 0 no change no change no change no change Si 5a Medium IGSCC Medium 0 2 2 -2.00E-11 -2.OOE-11 -2.OOE-12 -2.00E-12 Si 6a Medium None Low 33 0 -33 negligible negligible negligible negligible Si 7a Low None Low 13 0 -13 negligible negligible negligible negligible Si Total -1.79E-09 -1.79E-09 -5.40E-11 -5.40E-11 CS 2 High Tr Medium 0 1 1 -6.30E-09 -3.50E-09 -1.80E-10 -1.00E-10 CS 4 High None Low 1 3 2 -3.50E-10 -3.50E-10 -1.OOE-11 -1.OOE-11 CS 5a Medium Tr Medium 0 1 1 -1.80E-11 -1.002-11 -1.80E-12 -1.00E-12 CS 6a Medium None Low 7 0 -7 negligible negligible negligible negligible CS 7a Low None Low 0 0 0 no change no change no change no change CS Total -6.67E-09 .3.86E-09 -1.92E-10 -1.11E-10 Page A2-20 of 24

Table 3.6-1 (Cont'd)

Risk Impact Analysis Results Systemt 1( Category Consequence Failure Potential Inspections CDF Impact t 3) LERF Impactt 3 )

Rank DMs Rank SXIJ) RI-ISI Delta w/ POD [ who PODw POD/ w/o POD MS 6a Medium None Low 10 0 -10 negligible negligible negligible negligible MS Total l l l l __I_l ___negligible negligible negligible negligible FW 5a Medium TT Medium 1 1 0 -1.20E-11 no change -1.20E-12 no change FW 6a (3) Medium None (FAC) Low (High) 6 0 -6 negligible negligible negligible negligible FW 6a Medium None Low 2 0 -2 negligible negligible negligible negligible FW Total l _ __7___l_l -1.20E-11 negligible -1.20E-12 negligible CT 4 High None Low 3 2 -1 1.75E-10 1.75E-10 5.OOE-12 5.00E-12 CT 6a Medium None Low 0 0 0 no change no change no change no change CT 7a Low None Low 19 0 -19 negligible negligible negligible negligible CT Total l l lll_ l_l 1.75E-10 1.75E-10 5.00E-12 5.00E-12 AF 5a Medium TASCS Medium 1 1 0 -1.20E-11 no change -1.20E-12 no change AF 6a Medium None Low 12 0 -12 negligible negligible negligible negligible AF Total l 7 l l l l l -1.20E-11 negligible -1.20E-12 negligible Grand Total 7.43E.09 7.33E-08 2.05E.10 2.09E-09 Notes

1. Systems are described in Table 3.1.

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-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 A2-21 of 24

(.)

Table 5-1 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 Weid Section Xi EPRI TR-112657 Weld Section Xi EPRI TR-112657 Count Vol/Sur Sur Only Rl-ISI lOther(2) Count Vol/Sur Sur Onlyl Ri-ISI Other( 2 ) Count Vol/Sur -urOnly RI-ISI Other(2)

RC B-F 9 9 0 6 9 9 0 3 B-J 48 27 6 9 264 51 30 25 6 0 0 0 RHR B-J  ; 27 0 0 0 C-F-1 22 0 0 3 182 5 0 0 Si B-J 36 0 10 4 180 0 18 0 C-F-1 225 13 8 23 402 46 12 0 CS B-J 2 0 0O _ 32 0 8 4 _ _ 11 0 4 0 -

2C-F-i 5 s 1 0 0 191 7 0 0 MS C-F-2 103 10 0 0

. I I. I I -

FW C-F-2 6 1 0 1 73 8 0 0 CT C-F-1i 15 3 0 2 284 19 0 0 AF C-F-2 6 1 0 1 127 12 0 0 B-F 9 9 0 6 9 9 0 3 Total B-J 50 27 6 10 332 51 48 33 224 0 22 0 C-F-1 _ 267 17 8 28 1059 77 12 0 C-F-2 12 2 0 2 303 30 0 0 Notes

1. Systems are described in Table 3.1.

2. The column labeled 'Other" is generally used to identify plant augmented Inspection program locations credited per Section 3.6.5 of EPRI TR-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 HNP RI-ISI application. The 'Other" column has been retained in this table solely for uniformity purposes with the other RI-ISI application template submittals.

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Table 5-2 Inspection Location Selection Comparison Between ASME Section Xl Code and EPRI TR-112657 by Risk Category System(l)

Category Risk Rank JConsequence Rank DMs Failure Potential Rank Code C

Category Weld Count Section Xi EPRI TR-112657 ur Sur Only RI-ISI Other()

RC 2 High High TASCSTT Medium B-J 26 16 4 4 RC 2 High High TT, PWSCC Medium B-F 1 1 0 1I RC 2 High High TASCS Medium B-J 17 9 0 5 RC 2 High High TT Medium B-J 5 2 2 0 RC 2 High High PWSCC Medium B-F 8 8 0 5 B-F 9 9 0 ____

RC 4 Medium High None Low BF 2 9 0 25 B-J 264 51 30 25 ___

RC 7a Low Low None Low B-J 6 0 0 0 RHR 4 Medium High None Low C-F-i 16 0 0 2 ______

RHR 5a Medium Medium TASCS Medium C-F-I 6 0 0 1 X _

B-J 27 0 0 0 _ _

RHR 6a Low Medium None Low

. C-F-I 182 5 0 0 Si 4 Medium High None Low C-F-I 225 13 8 23 Si 5a Medium Medium Tr, IGSCC Medium B-J 13 0 7 2 Si 5a Medium Medium TT Medium B-J 3 0 0 0 Si 5a Medium Medium IGSCC Medium B-J 20 0 3 - 2 B-J 180 0 18 0 Si 6a Low Medium None Low C-F-1 207 33 0 0 Si 7a Low Low None Low C-F-1 195 13 12 0 CS 2 High High TT Medium B-J 2 0 0 1 CS 4 Medium High None Low B-J 24 0 8 3

___ __ _ _ _ C-F-i 5 1 0 0 _ _ _

CS 5a Medium Medium TT Medium B-J 8 0 0 1 CS 6a Low Medium None Low B-J 18 0 4 0 CS 7a LowLwNonLowC-F-i 183 7 0 0 Cs 7a Low Low None Low C-F-i 8 0 0 0 Page A2-23 of 24

Table 5-2 (Cont'd)

Inspection Location Selection Comparison Between ASME Section Xl Code and EPRI TR-112657 by Risk Category Risk Failure Potential Code Weld Section Xi EPRI TR-112657 system(') Category Rank Rank DMs Rank Category Count VolISur Sur Only RI-ISI Other( 2 )

MS 6a Low Medium None Low C-F-2 103 10 0 0 FW 5a Medium Medium Tr Medium C-F-2 6 1 0 1 FW 6a (3) Low (High) Medium None (FAC) Low (High) C-F-2 56 6 0 : 0 FW 6a Low Medium None Low C-F-2 17 2 0 0 CT 4 Medium High None Low C-F-1 15 3 0. 2 CT 6a Low Medium None Low C-F-1 49 0 0 0 '

CT 7a Low Low None Low C-F-1 235 19 0 0 AF 5a Medium Medium TASCS Medium C-F-2 6 1 0 1 -. .

AF 6a Low Medium None Low C-F-2 127 12 0 0 Notes

1. Systems are described in Table 3.1.

2. The column labeled 'Other" Is generally used to identify plant augmented Inspection program locations credited per Section 3.6.5 of EPRI TR-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 I weld population. This option was not applicable for the HNP RI-ISI application. The 'Other" column has been retained in this table solely for uniformity purposes with the other Ri-ISI application template submittals.

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