ML062850540

Request for Approval of Risk-Informed Inservice Inspection Program for Class 1 and 2 Piping American Society of Mechanical Engineers Code, Category B-F, B-J, C-F-1, and C-F-2 Piping Welds

ML062850540
Person / Time
Site:	Cook
Issue date:	09/29/2006
From:	Jensen J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP:NRC:6055-09
Download: ML062850540 (38)

Text

Indiana Michigan Power INDIANA Cook Nuclear Plant MICHIGAN PFOMWR

• One Cook Place Bridgman, Ml 49106 AEP~com A unitof American Electric Power September 29, 2006 AEP:NRC:6055-09 10 CFR 50.55a(3)(i)

Docket Nos. 50-315 50-316 U. S. Nuclear Regulator Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001 Donald C. Cook Nuclear Plant Units I and 2 REQUEST FOR APPROVAL OF RISK-INFORMED INSERVICE INSPECTION PROGRAM FOR CLASS 1 AND 2 PIPING AMERICAN SOCIETY OF MECHANICAL ENGINEERS CODE, CATEGORY B-F, B-J, C-F-I, AND C-F-2 PIPING WELDS Pursuant to the requirements of 10 CFR 50.55a(3)(i), Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Units I and 2, is proposing an alternative to the requirements of the 1989 Edition of the American Society of Mechanical Engineers Code,Section XI, in order to implement a Risk-Informed Inservice Inspection (RI-ISI) Program for Class 1, Code Category B-F, B-J, C-F-I, and C-F-2 piping welds. The program, which is described in Attachment I to this letter, has been developed in accordance with Code Case N-716, "Alternative Piping Classification and Examination Requirements."

I&M plans to implement the proposed RI-ISI Program during the third period of the third 10-year inservice inspection interval that began on July 1, 1996. I&M requests approval of the RI-ISI Program by September 30, 2007, to facilitate planning for the remainder of the Third Ten-Year Inservice Inspection Interval.

In making this submittal, I&M is volunteering CNP as the pressurized water reactor (PWR) pilot plant for the use of Code Case N-716 as the basis for an RI-ISI program.

The provisions of 10 CFR 170.11(a)(1)(ii) and (iii) state that no application fees, license fees, renewal fees, inspection fees, or special project fees shall be required to assist the Nuclear Regulatory Commission (NRC) in developing a rule, regulatory guide, policy state, generic letter, or bulletin; or as a means of exchanging information between industry organizations and the NRC for the specific purpose of supporting the NRC's generic regulatory improvements or efforts. In that the I&M development and NRC review of the licensing actions required for the implementation of a PWR RI-ISI program based on Code Case N-716 would satisfy these provisions, I&M will be requesting a waiver of the associated NRC fees in accordance with 10 CFR 170.5 in a separate letter.

AOL/iw

U. S. Nuclear Regulatory Commission AEP:NRC:6055-09 Page 2 Commitments made in this letter are identified in Attachment 2. Should you have any questions, please contact Ms. Susan D. Simpson, Regulatory Affairs Manager, at (269) 466-2428.

Sincerely, Services Vice President Attachments: 1. Donald C. Cook Nuclear Plant Units 1 and 2, Risk-Informed/Safety-Based Inservice Inspection Program Plan.

2. Regulatory Commitment.

RGV/jen c: R. Aben - Department of Labor and Economic Growth J. L. Caldwell - NRC Region III K. D. Curry - AEP Ft. Wayne, w/o attachments J. T. King - MPSC MDEQ - WHMD/RPMWS NRC Resident Inspector P. S. Tam - NRC Washington, DC

Attachment 1 to AEP:NRC:6055-09 Donald C. Cook Nuclear Plant Units 1 and 2 RISK-INFORMED/SAFETY-BASED INSERVICE INSPECTION PROGRAM PLAN Table of Contents

1. Introduction 1.1 Relation to Nuclear Regulatory Commission (NRC) Regulatory Guides 1.174 and 1.178 1.2 Probabilistic Risk Assessment (PRA) Quality

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

3. Risk-Informed/Safety-Based ISI Process 3.1 Safety Significance Determination 3.2 Failure Potential Assessment 3.3 Element and Non-Destructive Examination (NDE) Selection 3.3.1 Additional Examinations 3.3.2 Program Relief Requests 3.4 Risk Impact Assessment 3.4.1 Quantitative Analysis 3.4.2 Defense-in-Depth

4. Implementation and Monitoring Program

5. Proposed ISI Program Plan Change

6. References/Documentation to AEP:NRC:6055-09 Page 2

1. INTRODUCTION The Donald C. Cook Nuclear Plant (CNP) is currently in the third 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. Indiana Michigan Power Company (I&M), the licensee for CNP, plans to start implementing a risk-informed/safety-based inservice inspection (RISB) program during the third inspection period of the current interval. Initial RISB Program implementation is planned for each unit as indicated below. The ASME Section XI Code of Record for the third ISI interval at CNP is the 1989 Edition.

Unit Refueling Outage Number Scheduled Start 1 Unit 1, Cycle 22 Spring 2008 2 Unit 2, Cycle 17 Fall 2007 The objective of this submittal is to request the use of the RISB process for the inservice inspection of Class 1 and 2 piping. The RISB process used in this submittal is based upon Reference 1, which is founded in large part on the Risk-Informed ISI (RI-ISI) process as described in Reference 2.

1.1 Relation to Nuclear Regulatory Commission (NRC) Regulatory Guides 1.174 and 1.178 As a risk-informed application, this submittal meets the intent and principles of References 3 and 4. Further information is provided in Section 3.6.2 relative to defense-in-depth.

1.2 Probabilistic Risk Assessment (PRA) Quality I&M has exhibited a long-term commitment to using, maintaining, and updating the CNP PRA model. In 1992, I&M submitted responses, including a Level 3 internal events PRA, seismic PRA, and a fire PRA, to fulfill the requirements of NRC Generic Letter 88-20. In 1995, I&M submitted extensive revisions to the human reliability analysis, seismic, and fire models. Further PRA model updates to address plant modifications and to update data were completed in 1996 and 1997. In June 2001, I&M completed a project to update and make other improvements to the existing version of the PRA. The overall purpose of this project was to enhance the usage of the PRA model to support compliance with 10 CFR 50.65(a)(4) for management of risk during maintenance activities, and to support the new risk-informed, performance-based regulatory environment. This project included:

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" Updating the PRA model to include new plant specific data, making necessary model changes because of procedure and/or design changes, updating the treatment of common-cause failures, and removing unnecessary or unwarranted conservatisms and simplifications;

" Adding a large early release frequency (LERF) model to the PRA model; and

" Developing a separate Unit 2 model.

In September 2001, the updated PRA model received a certification review in accordance with the Westinghouse Owner's Group (WOG) certification process. This review led to a number of Facts and Observations (F&Os), including three "A" Level significance F&Os and 24 "B" Level significance F&Os. The WOG Certification process assigns "A" Level significance to F&Os that are considered extremely important and necessary to address to assure the technical adequacy or quality of the PRA model, while "B" Level significance is assigned to F&Os that are considered important and necessary to address, but may be deferred until the next PRA model update.

Following receipt of the draft WOG certification report, I&M undertook a model update that addressed all of the "A" and "B" Level F&Os, with the exception of an "A" Level F&O that concerned internal flooding. The goal of these updates was to assure that the F&Os were addressed sufficiently to meet the criteria identified for ASME PRA Standard Quality Category 2. Implementation of these changes to the PRA notebooks was completed in October 2003. Quantification of the revisions was completed in April 2004.

Subsequent to the latest update effort, I&M had the F&O resolutions reviewed and validated as satisfactory by an independent contractor. This contractor also performed a gap assessment of the updated model compared to Regulatory Guide 1.200. Internal flooding was recently addressed (2006) to complete the effort to address all WOG certification Level A and B F&Os.

Since the completion of the major updates, there have been two focused-scope updates related to increased Station Blackout mitigation capability. Specifically, in August 2005, the CNP PRA model was updated to include the addition of Supplemental Diesel Generators, which provide a non-safety-related, on-site back-up power source for the safety-related Emergency Diesel Generators. In June 2006, the CNP PRA model was updated to include a change in procedural guidance for beyond-licensing-basis Station Blackout that directs usage of the Charging system inter-unit cross-tie.

The CNP Reference 1 application uses bounding high values relative to CNP's latest PRA model. The base case Core Damage Frequency (CDF) is 2.1E-05/year, and the base case LERF is 4.2E-06/year.

to AEP:NRC:6055-09 Page 4 Based on the foregoing, it is judged that the current PRA model, used in the RIS_B evaluation, has an acceptable quality to support this application.

2. PROPOSED ALTERNATIVE TO CURRENT ISI PROGRAMS 2.1 ASME Section XI ASME Section XI Examination Categories B-F, B-J, C-F-I, and C-F-2 currently contain the requirements for the nondestructive examination (NDE) of Class 1 and 2 piping components. The alternative RISB Program for piping is described in Reference 1. The RISB Program will be substituted for the current program for Class 1 and 2 piping (Examination Categories B-F, B-J, C-F-i, and C-F-2) in accordance with 10 CFR 50.55a(a)(3)(i) by alternatively providing an acceptable level of quality and safety. Other non-related portions of the ASME Section XI Code will be unaffected.

2.2 Augmented Programs The plant augmented inspection programs listed below were considered during the RISB application. It should be noted that this section documents only those plant augmented inspection programs that address common piping with the RISB application scope (i.e., Class 1 and 2 piping).

• The plant augmented inspection program for flow accelerated corrosion per Generic Letter 89-08 is relied upon to manage this damage mechanism, but is not otherwise affected or changed by the RISB Program.

• A plant augmented inspection program is being implemented at CNP in response to Reference 5. The requirements of MRP-139 will be used for the inspection and management of primary water stress corrosion cracking (PWSCC) susceptible welds and will supplement the RISB Program selection process. The RIS_B Program will not be used to eliminate any Reference 5 requirements.

• I&M is in the process of evaluating MRP-146, "Materials Reliability Program:

Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines," and these results will be incorporated into the RISB Program if warranted.

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3. RISK-INFORMED/SAFETY-BASED ISI PROCESS The process used to develop the RISB Program conformed to the methodology described in Reference 1 and consisted of the following steps:

• Safety Significance Determination

• Failure Potential Assessment

• Element and NDE Selection

• Risk Impact Assessment

• Implementation Program

• Feedback Loop 3.1 Safety Significance Determination The systems assessed in the RISB Program are provided in Tables 3.1-1 and 3.1-2 for Units 1 and 2, respectively. The piping and instrumentation diagrams and additional plant information, including the existing plant ISI Program, were used to define the piping system boundaries.

Per Reference 1 requirements, piping welds are assigned safety significance categories that are used to determine the treatment requirements. High safety significant (HSS) welds are determined in accordance with the requirements below. Low safety significant (LSS) welds shall include all other Class 2, 3, or Non-Class welds.

(1) Class 1 portions of the reactor coolant pressure boundary (RCPB), except as provided in (c)(2)(i) and (c)(2)(ii) of 10 CFR 50.55a; (2) applicable portions of the shutdown cooling pressure boundary function shall be included. That is, Class 1 and 2 welds of systems or portions of systems needed to utilize the normal shutdown cooling flowpath either:

(i) as part of the RCPB from the reactor pressure vessel (RPV) to the second isolation valve (i.e., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds, or (ii) other systems or portions of systems from the RPV to the second isolation valve (i.e., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds; to AEP:NRC:6055-09 Page 6 (3) that portion of the Class 2 feedwater system [> nominal pipe size (NPS) 4 inches] of Pressurized Water Reactors (PWRs) from the steam generator to the outer containment isolation valve; (4) piping within the break exclusion region [> NPS 4 inches] for high energy piping systems as defined by the Owner; and (5) any piping segment whose contribution to core damage frequency is greater than 1E-06 based upon a plant-specific PRA of pressure boundary failures (e.g., pipe whip, jet impingement, spray, inventory losses). This may include Class 3 or Non-Class piping.

3.2 Failure Potential Assessment Failure potential estimates were generated utilizing industry failure history, plant specific failure history, and other relevant information. These failure estimates were determined using the guidance provided in Reference 2, with the exception of the deviation discussed below.

Tables 3.2-1 and 3.2-2 summarize the failure potential assessment by system for each degradation mechanism that was identified as potentially operative for Units 1 and 2, respectively.

A deviation to Reference 2 has been implemented in the failure potential assessment for CNP. Table 3-16 of Reference 2 contains criteria for assessing the potential for thermal stratification, cycling, and striping (TASCS). Key attributes for horizontal or slightly sloped piping greater than 1-inch NPS include:

1. 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 to AEP:NRC:6055-09 Page 7

> temperature differential (AT) > 50'F; AND

> Richardson Number > 4 (this value predicts the potential buoyancy of a stratifliedflow).

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.

1. 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.

2. Low flow TASCS In some situations, the transient startup of a system (e.g., residual heat removal system 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 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 to AEP:NRC:6055-09 Page 8 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 will govern.

3. 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.

4. 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 by References 6 and 7. The methodology used in the CNP RISB application for assessing TASCS potential conforms to these updated criteria. Final materials reliability program guidance on the subject of TASCS will be incorporated into the CNP RISB application if warranted. It should be noted that the NRC has granted approval (References 8 and 9) for RI-ISI relief requests incorporating these TASCS criteria at several facilities, including Comanche Peak and South Texas Project.

3.3 Element and NDE Selection Reference 1 provides criteria for identifying the number and location of required examinations. Ten percent of the high safety significant welds shall be selected for examination as follows:

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

(a) A minimum of 25 percent of the population identified as susceptible to each degradation mechanism and degradation mechanism combination shall be selected.

(b) If the examinations selected above exceed ten percent of the total number of high safety significant welds, the examinations may be reduced by to AEP:NRC:6055-09 Page 9 prorating among each degradation mechanism and degradation mechanism combination, to the extent practical, such that at least ten percent of the high safety significant population is inspected.

(c) If the examinations selected above are not at least ten percent of the high safety significant weld population, additional welds shall be selected so that the total number selected for examination is at least ten percent. The additional welds will be selected in accordance with Reference 1.

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

(3) A minimum of ten percent of the welds in that portion of the RCPB that lies outside containment shall be selected.

(4) A minimum of ten percent of the welds within the break exclusion region shall be selected.

In contrast to a number of RI-ISI Program applications where the percentage of Class 1 piping locations selected for examination has fallen substantially below ten percent, Reference 1 mandates that ten percent be chosen. A brief summary is provided below, and the results of the selections are presented in Tables 3.3-1 and 3.3-2 for Units I and 2, respectively. Section 4 of Reference 2 was used as guidance in determining the examination requirements for these locations.

Class I Piping Weldstt) Class 2 Piping Welds (2) All Piping Welds ()

Unt Total Selected Total Selected Total Selected 1 1196 124 1730 22 2926 146 2 1217 126 1628 20 2845 146 Notes

1. Includes all Category B-F and B-J locations. All 1196 Unit 1 Class I piping weld locations and 1217 Unit 2 Class I piping weld locations are HSS.

2. Includes all Category C-F-I and C-F-2 locations. Of the 1730 Unit I Class 2 piping weld locations, 240 are HSS and the remaining 1490 are LSS. Of the 1628 Unit 2 Class 2 piping weld locations, 228 are HSS and the remaining 1400 are LSS.

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

to AEP:NRC:6055-09 Page 10 3.3.1 Additional Examinations The RISB Program in all cases will determine through an engineering evaluation the 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 causal conditions. Additional examinations will be performed on those elements with the same causal conditions or degradation mechanisms. The additional examinations will include high safety significant elements up to a number equivalent to the number of elements required to be inspected 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 causal conditions.

3.3.2 Program Relief Requests An attempt has been made to select RISB 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 Reference 2 will be followed.

Relief requests ISIR-005 and ISIR-006, which are currently in CNP's ISI program will be withdrawn following NRC approval of the RISB Program. These relief requests pertain to the surface and volumetric examination of feedwater and main steam pipe-to-flued head welds. The pipe-to-flued head welds in the feedwater system are included in the scope that is designated high safety significant, but have not been selected for examination. The main steam system in its entirety is designated low safety significant and is not subject to NDE.

to AEP:NRC:6055-09 Page I11 3.4 Risk Impact Assessment The RISB Program has been conducted in accordance with Reference 3 and the requirements of Reference 1, and the risk from implementation of this program is expected to remain neutral or decrease when compared to that estimated from current requirements.

This evaluation categorized segments as HSS or LSS in accordance with Reference 1, and then determined what inspection changes are proposed for each system. The changes include changing the number and location of inspections and, in many cases, improving the effectiveness of the inspection to account for the findings of the RISB 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.

3.4.1 Quantitative Analysis Reference 1 has adopted the Electric Power Research Institute (EPRI) RI-ISI process for risk impact analyses whereby limits are imposed to ensure that the change in risk of implementing the RISB Program meets the requirements of References 3 and 4. The EPRI criterion requires that the cumulative change in CDF and LERF be less than 1E-07 and 1E-08 per year per system, respectively.

I&M has conducted a risk impact analysis per the requirements of Section 5 of Reference I that is consistent with the "Simplified Risk Quantification Method" described in Section 3.7 of Reference 2. The analysis estimates the net change in risk due to the positive and negative influence of adding and removing locations from the inspection program.

The conditional core damage probability (CCDP) and conditional large early release probability (CLERP) values used to assess risk impact were estimated based on pipe break location. Based on these estimated values, a corresponding consequence rank was assigned per the requirements of Reference 2 and upper bound threshold values were used as provided on the following page. Consistent with Reference 2, the upper bound for all break locations that fall within the high consequence rank range was based on the highest CCDP value obtained (i.e.,

large break loss-of coolant accident (LOCA) for CNP).

to AEP:NRC:6055-09 Page 12 CCDP and CLERP Values Based on Break Location Break Location Estimated Consequence Upper Bound CCDP CLERP Rank CCDP CLERP LOCA 3. 00E-02 3. OOE-03 HIGH 3. OOE-02 3. 00E-03 RCPB pipe breaks that result in a loss of coolant accident (LOCA) - The highest CCDP for Large Break LOCA (LBLOCA) was used as the basis for estimating (0.1 margin used for CLERP)

ILOCA 1. 20E-04 1. 20E-05 1llGH1 3. 00E-02 3. OOE-03 RCPB pipe breaks that result in an isolable LOCA (ILOCA) - Calculated based on LBLOCA CCDP of 3E-2 and motor operated valve failure to close on demand rate of 4E-3 (0.1 margin used for CLERP)

PLOCA 6. OOE-05 6. OOE-06 MEDIUM L.OOE-04 L.00E-05 RCPB pipe breaks that result in a potential LOCA (PLOCA) - Calculated based on LBLOCA CCDP of 3E-2 and check valve (CV) rupture or failure to close rate of 2E-3 (0.1 margin used for CLERP)

PILOCA - OC 2.OOE-03 2.OOE-03 1IGH1 3. 00E-02 3. OOE-03 RCPB pipe breaks that result in a potential isolable LOCA outside containment (PILOCA - OC) -

Calculated based on a CCDP and CLERP of 1.0 for LOCA outside containment and a CV rupture rate of 2E-3 [isolation not credited]

PILOCA - IC 6. OOE-05 6. OOE-06 MEDIUM L.OOE-04 L.OOE-05 RCPB pipe breaks that result in a potential isolable LOCA inside containment (PILOCA - IC) - Calculated based on LBLOCA CCDP of 3E-2 and a CV rupture rate of 2E-3 (0.1 margin used for CLERP) [isolation not credited]

Class 2 SDC - IC L.OOE-04 1.OOE-05 MEDIUM L.OOE-04 1.00E-05 Non RCPB pipe breaks that occur in Class 2 shutdown cooling piping inside containment (Class 2 SDC -

IC) - Estimated based on a loss of shutdown cooling during mid-loop operation (0.1 margin used for CLERP)

Class 2 FWU - OC 1.OOE-03 1. 50E-04 HIGH 3. OOE-02 3. OOE-03 Non RCPB pipe breaks that occur in Class 2 feedwater piping unisolable from the steam generator outside containment (Class 2 FWU - OC) - The CCDP for steam line break outside containment (SLBO) for piping downstream of the main steam isolation valve (MSIV) was used as the basis for estimating (0.15 margin used for CLERP)

Class 2 FWU - IC L.OOE-03 1. 50E-04 IGH1 3. OOE-02 3. OOE-03 Non RCPB pipe breaks that occur in Class 2 feedwater piping unisolable from the steam generat6r inside containment (Class 2 FWU - IC) - The CCDP for steam line break inside containment was used as the basis for estimating (0.15 margin used for CLERP)

Class 2 F*VI - OC L.OOE-03 1. 50E-04 HIGH 3. OOE-02 3. OOE-03 Non RCPB pipe breaks that occur in Class 2 feedwater piping isolable from the steam generator outside containment (Class 2 FWI - OC) - The CCDP for SLBO for piping downstream of the MSIV was used as the basis for estimating (0.15 margin used for CLERP)

Class 2 LSS 1.00E-04 1.00E-05 MEDIUM 1.00E-04 1.00E-05 Non RCPB pipe breaks that occur in all other Class 2 system piping designated as low safety significant (Class 2 LSS) - Estimated based on upper bound for Medium Consequence to AEP:NRC:6055-09 Page 13 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 lE-08. Piping locations identified as medium failure potential have a likelihood of 20xo. These PBF likelihoods are consistent with those listed in References 9 and 14 of Reference 2. 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 RISB approach.

Tables 3.4-1 and 3.4-2 present summaries of the RISB Program versus 1989 ASME Section XI Code Edition program requirements on a per system basis.

As indicated on the following page, this evaluation has demonstrated that unacceptable risk impacts will not occur from implementation of the RIS_B Program, and satisfies the acceptance criteria of References 1 and 3.

to AEP:NRC:6055-09 Page 14 Unit 1 Risk Impact Results ARiskcDF ARiskLERr System(')

w/ POD w/o POD w/ POD [ w/o POD RC -2.21E-08 -1.65E-09 -2.21E-09 -1.65E-10 CS -3.77E-08 -2.09E-08 -3.77E-09 -2.09E-09 RH -2.70E-I I -2.70E- I1 -2.70E-12 -2.70E- 12 SI -4.15E-lI -4.15E- 1 -4.15E-12 -4.15E-12 FW -2.70E-09 8.1OE-09 -2.70E-10 8.10E-10 MS 1.60E-10 1.60E-10 1.60E- I I 1.60E-1I CTS 1.40E-10 1.40E-10 1.40E-1I 1.40E-1I Total -6.22E-08 -1.42E-08 -6.22E-09 -1.42E-09 Note I Systems are defined on Page 19 Unit 2 Risk Impact Results System(i) ARiskCDF ARiskLERF w/POD w/o POD w/ POD J w/o POD RC -2.22E-08 -1.80E-09 -2.22E-09 -1.80E-10 CS -3.77E-08 -2.09E-08 -3.77E-09 -2.09E-09 RH -4.70E- I1 -4.70E-I I -4.70E-12 -4.70E- 12 SI -4.03E-10 -4.03E-10 -4.03E-I1 -4.03E- I I FW -9.45E-09 -3.45E-09 -9.45E-10 -3.45E-10 MS 1.70E- 10 1.70E-10 1.70E-I 1 1.70E-I I CTS 7.OOE- I1 7.OOE-I I 7.OOE-12 7.OOE- 12 Total -6.95E-08 -2.63E-08 -6.95E-09 -2.63E-09 Note

1. Systems are defined on Page 19.

Attachment I to AEP:NRC:6055-09 Page 15 3.4.2 Defense-in-Depth The intent of the inspections mandated by ASME Section XI 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, Revision 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.

References 1 and 2 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 leaks or ruptures is increased. Secondly, a generic assessment of high consequence sites has been determined by Reference 1, supplemented by plant-specific evaluations, thereby requiring a minimum threshold of inspection for important piping whose failure would result in a LOCA or break exclusion region (BER) break. Finally, Reference 1 requires that any piping on a plant-specific basis that has a contribution to CDF of greater than 1E-06 be included in the scope of the application. No such piping was identified at CNP.

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 safety significance.

4. IMPLEMENTATION AND MONITORING PROGRAM Upon approval of the RISB Program, procedures that comply with the guidelines described in Reference 2 will be prepared to implement and monitor the program. The new program will be integrated into the third inservice inspection interval. 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 XI program implementing procedures will be retained and modified to address the RISB process, as appropriate.

Attachment 1 to AEP:NRC:6055-09 Page 16 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 RISB 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, this review will be conducted 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 RIS B Program and ASME Section XI 1989 Code Edition program requirements for in-scope piping is provided in Tables 5-1 and 5-2 for Units 1 and 2, respectively.

CNP plans to start implementing the RISB Program during the third inspection period. Initial RISB Program implementation is planned for each unit as indicated below. Upon implementation, inspection locations selected per the RISB process will replace those formerly selected per ASME Section XI criteria. The table below indicates the percentage of piping weld examinations required by ASME Section XI that have been completed thus far in the second ISI interval for Examination Categories B-F, B-J, C-F-i, and C-F-2 through the end of the second period. The table also indicates the percentage of inspection locations selected for examination per the RISB process that will be examined in the third period to ensure the performance of 100% of the required examinations during the current ten-year ISI interval as discussed in Relief Request ISIR-19, Reference 10.

Attachment 1 to AEP:NRC:6055-09 Page 17 Unit Refueling Outage Examination Percentages Number Scheduled Start ASME Section XI RISB Program I U1 C22 Spring 2008 34% 66%

2 U2 C17 Fall 2007 34% 66%

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

REFERENCES

1. ASME Code Case N-716, "Alternative Piping Classification and Examination Requirements,Section XI Division 1," dated April 19, 2006.

2. EPRI TR-1 12657, Revision B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure," dated December 1999.

3. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis," dated November 2002.

4. Regulatory Guide 1.178, "An Approach for Plant-Specific Risk-Informed Decisionmaking for Inservice Inspection of Piping," dated September 2003.

5. MRP-139, "Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guidelines," dated August 2005.

6. Letter from Pat O'Regan, EPRI, to Brian W. Sheron, NRC, "Extension of Risk-Informed Inservice Inspection (RI-ISI) Methodology," Accession Number ML010650169, dated February 28, 2001.

7. Letter from Pat O'Regan, EPRI, to Brian W. Sheron, NRC, "Extension of Risk-Informed Inservice Inspection (RI-ISI) Methodology," Accession Number ML011070238, dated March 28, 2001.

8. Letter from Robert A. Gramm, NRC, to C. Lance Terry, TXU Electric, "Comanche Peak Steam Electric Station (CPSES), Units 1 and 2 - Approval of Relief Request for Application of Risk-Informed Inservice Inspection Program for American Society of Mechanical Engineers Boiler and Pressure Vessel Code Class 1 and 2 Piping (TAC Nos.

MB1201 and MB1202)," Accession Number ML012710112, dated September 28, 2001.

to AEP:NRC:6055-09 Page 18

9. Letter from Robert A. Gramm, NRC, to William T. Cottle, STP Nuclear Operating Company, "Approval of Relief Request for Application of Risk-Informed Inservice Inspection program for American Society of Mechanical Engineers Boiler and Pressure Vessel Code Class 1 and 2 Piping for South Texas Project, Units land 2 (TAC Nos.

MB1277 and MB1278)," Accession Number ML020390041, dated March 5, 2002.

10. Letter from Daniel P. Fadel, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, Proposed Alternative to the American Society of Mechanical Engineers Code,Section XI Weld Inspection Requirements," AEP:NRC:5055-10, Accession Number ML052780450, dated September 22, 2005.

to AEP:NRC:6055-09 Page 19 The following provides the definition of the terms used in Table 3.1-1 thru Table 5-2.

BER Break Exclusion Region CS Charging System CC Crevice Corrosion CDF Core Damage Frequency CTS Containment Spray System DMs Degradation Mechanisms ECSCC External Chloride Stress Corrosion Cracking E-C Erosion-Cavitation FAC *Flow-Accelerated Corrosion FW Feedwater System HSS High Safety Significant IGSCC Intergranular stress corrosion cracking IFIV Inside First Isolation Valve LERF Large Early Release Frequency LSS Low Safety Significant MIC Microbiologically-Influenced Corrosion MS Main Steam System OC Outside Containment PIT Pitting POD Probability of Detection PWR: FW Pressurized Water Reactor: Feedwater PWSCC Primary Water Stress Corrosion Cracking RC Reactor Coolant System RCPB Reactor Coolant Pressure Boundary RH Residual Heat Removal System SDC Shutdown Cooling Piping SI Safety Injection System TASCS Thermal Stratification, Cycling, and Striping TGSCC Transgranular Stress Corrosion Cracking TT Thermal Transients Vol Volume Sur Surface to AEP:NRC:6055-09 Page 20 Table 3.1-1 N-716 Safety Significance Determination for Unit 1 1 N-716 Safety Significance Determination Safety Significance RCPB SDC PWR: FW BER ICDF> 1E.-6 High Low RC 662 " "

CS 70 V 239 "

RH 22 V V ,

26 V "

282 "

SI 51 V " V 391 V V 586 "

FW 214 V "

MS 218 V CTS 165 V

SUMMARY

RESULTS FOR ALL SYSTEMS 73 V V "

1123 V V 26 V V 214 V V 1490 "

TOTALS 2926 1436 1490 Note

1. Systems are defined on Page 19.

to AEP:NRC:6055-09 Page 21 Table 3.1-2 N-716 Safety Significance Determination for Unit 2 System Description~') 1Weld Count 1~RP N-716 Safety Significance Determination E _____ _____Safety Significance RCPB SDC PWR: FW BER CDFh> I" ligh Low RC 669 1 CS 64 " "

213 1 Rtt 27 "'

28 1 273 "

SI 56 1 1 1 401 le 545 "

FW 200 1 I MS 211 CTS 158 "

SUMMARY

RESULTS FOR ALL SYSTEMS 83 "

1134 1 I 28 1" V" 200 "

1400 1 TOTALS 2845 1445 1400 Note

1. Systems are defined on Page 19.

Attachment I to AEP:NRC:6055-09 Page 22 Table 3.2-1 Failure Potential Assessment Summary for Unit 1 Sstemtt ) Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive TASCS TT IGSCC J TGSCC 1 ECSCC I PWSCC MIC PIT CC E-C FAC RC

CS(2)

SI(2)"

FW "

MS(2)

CTS(2)

Notes

1. Systems are defined on Page 19.

2. A degradation mechanism assessment was not performed on low safety significant piping segments. This includes the MS and CTS systems in their entirety, as well as portions of the CS, RIt and SI systems.

to AEP:NRC:6055-09 Page 23 Table 3.2-2 Failure Potential Assessment Summary for Unit 2 Stress Corrosion Cracking Localized Corrosion Flow Sensitive Ssel)

TASCS F Thermal Fatigue TT IGSCC TGSCC ECSCC PWSCC

MIC PIT CC I" E-C E"

FAC RC CS(2)

• RH (2)

Si(2)

FW MS(2)

CTS(2)

Notes

1. Systems are defined on Page 19.

2. A degradation mechanism assessment was not performed on low safety significant piping segments. This includes the MS and CTS systems in their entirety, as well as portions of the CS, RH and SI systems.

Attachment I to AEP:NRC:6055-09 Page 24 Table 3.3-1 N-716 Element Selections for Unit 1 Systemo) Weld Count N-71611 Selection Considerations 1Selectionst 2

)

i1SS LSS DMs RCPB""v RCPBOC BER RC 6 TASCS, TT 2 RC I TT, PWSCC I RC 12 TASCS / 3 RC 2 Tr T/ I RC 9 PWSCC / 9 RC 621 None " 51 RC I1 None 0 CS 28 TT " 7 CS 4 TI' 0 CS 16 None V" 0 CS 22 None 0 CS 239 RH 3 None " 2 RH 19 None 3 RH 26 None NA NA 0 RH 282 SI 47 IGSCC 14 SI 32 None " 8 SI 9 None 2 SI 354 None 21 SI 586 FW 8 TASCS NA NA 2 FW 206 None NA NA 20 MS 218 CTS 165

SUMMARY

6 TASCS, TI " 2 RESULTS I TI, PWSCC I FOR ALL SYSTEMS 12 TASCS 3" 8 TASCS NA NA 2 30 TI " 8 4 TI 0 47 IGSCC 14 9 PWSCC ' 9 to AEP:NRC:6055-09 Page 25 Table 3.3-1 (Cont'd)

N-716 Element Selections for Unit 1 SYstem~1t )1I Weld Count N-716 Selection Considerations rISelections* i 2

• IISS LSS DMs J RCPB"'t RCPBoc BER

SUMMARY

672 None " 61 RESULTS 9 None " 2 FOR ALL SYSTEMS 406 None 24 (CONT'D) 232 None NA NA 20 1490 TOTALS 1436 1490 146 Notes

1. Systems are defined on Page 19.

2. For RH and SI, Code Case N-716 Requirement 4(c) could not be inet. At least 25 percent of the welds located between the first isolation valve (i.e., the isolation valve closest to the RPV) and the RPV were selected.

to AEP:NRC:6055-09 Page 26 Table 3.3-2 N-716 Element Selections for Unit 2 SystemO' 1 Weld Count N-71611Selection Considerations Selections~2

• IISS LSS DNIs RCPB'"v RCPBoc BER RC 8 TASCS, Tr " 2 RC I TIT, PWSCC I RC 10 TASCS " 3 RC 2 TT I RC 5 PWSCC " 5 RC 632 None " 55 RC 11 None 0 CS 29 "IT T 7 CS 5 "T 0 CS 14 None " 0 CS 16 None 0 CS 213 RHI 3 None " 2 RH 24 None 4 RH 28 None NA NA 0 RH 273 SI 43 IGSCC 12 SI 32 None 1 8 SI 9 None 1 2 SI 373 None 24 SI 545 FW 8 TASCS NA NA 2 FW 192 None NA NA 18 MS 211 CTS 158

SUMMARY

8 TASCS, "T 2 RESULTS I "T, PWSCC I FOR ALL SYSTEMS 10 TASCS " 3 8 TASCS NA NA 2 31 UT 8 5 TU 0 43 IGSCC 12 5 PWSCC " 5 to AEP:NRC:6055-09 Page 27 Table 3.3-2 (Cont'd)

N-716 Element Selections for Unit 2 Systemtt T Weld r1 Count N-716 Selection Considerations Selectionst 21 IISS LSS DMs RCPB'vI RCPBoc BER

SUMMARY

681 None V" 65 RESULTS 9 None " 2 FOR ALL SYSTEMS 424 None 28 (CONT'D) 220 None NA NA 18 1400 TOTAL 1445 1400 146 Notes

1. Systems are defined on Page 19.

2. For RH and SI, Code Case N-716 Requirement 4(c) could not be met. At least 25 percent of the welds located between the first isolation valve (i.e., the isolation valve closest to the RPV) and the RPV were selected.

to AEP:NRC:6055-09 Page 28 Table 3.4-1 Risk Impact Analysis Results for Unit 1 System~') Safety Break Location Failure Potential Inspections CDF Impact LERF Impact Significance DNMs Rank SX1( j RISB Delta w/POD w/o POD w/ POD I w/o POD RC High LOCA TASCS, Tr Medium 3 2 -1 -5.40E-09 3.00E-09 -5.4013-10 3.0013-10 RC Hligh LOCA TT, PWSCC Medium 1 1 0 0.00E+00 0.0013+00 0.0013+00 0.00E+00 RC High LOCA TASCS Medium 2 3 1 -1.26E-08 -3.00E-09 -1.26E-09 -3.00E-10 RC High LOCA TT Medium 0 1 1 -5.40E-09 -3.0013-09 -5.40E-10 -3.0013-10 RC High LOCA PWSCC Medium 9 9 0 0.00E+00 0.0013+00 0.0013+00 0.0013+00 RC High LOCA None Low 60. 51 -9 1.35E-09 1.35E-09 1.35E-10 1.35E-10 RC High PLOCA None Low 0 0 0 0.0013+00 0.0013+00 0.00E+00 0.00E+00 RC TOTAL -2.21E-08 -1.65E-09 -2.21 E-09 -1.65E-10 CS High LOCA TT Medium 0 7 7 -3.78E-08 -2.10E-08 -3.78E-09 -2.1OE-09 CS High ILOCA TT Medium 0 0 0 0.0013+00 0.00E+00 0.00E+00 0.00E+00 CS High PLOCA TT Medium 0 0 0 0.00E+00 0.0013+00 0.0013+00 0.003E+00 CS High LOCA None Low 0 0 0 0.00E+00 0.0013+00 0.00E+00 0.00E+00 CS Iligh PLOCA None Low 0 0 0 0.0013+00 0.0013+00 0.0013+00 0.OOE+00 CS Low Class 2 LSS NIA Assume Medium II 0 -11 1.1013-10 1.10E-10 1.10E-I1 1.1013-11 CS TOTAL -3.77E-08 -2.09E-08 -3.77E-09 -2.09E-09 RII High LOCA None Low 1 2 1 -1.5013-10 -1.5013-10 -1.5013- 11 -1.5013- 11 RH High PLOCA None Low 5 3 -2 1.0013-12 1.0013-12 1.0013-13 1.0013-13 RH High Class 2 SDC-IC None Low 4 0 -4 2.0013-12 2.0013-12 2.0013-13 2.0013-13 RH Low Class 2 LSS NIA Assume Medium 12 0 -12 1.20E-10 1.20E-10 1.20E-11 1.20E-11 Ru TOTAL -2.70E-1 I -2.70E-11 -2.70E-12 -2.70E-12 to AEP:NRC:6055-09 Page 29 Table 3.4-1 (Cont'd)

Risk Impact Analysis Results for Unit I s Safety Break Location Failure Potential Inspections CDF Impact LERF Impact System Significance DMs Rank SXI(2 ) RISB Delta w/ POD w/o POD w/ POD w/o POD SI High PLOCA IGSCC Medium 13 14 1 -I.00E- 11 -1.00E- I1 -1.00E-12 -1.00E-12 SI High LOCA None Low 5 8 3 -4.50E-10 -4.50E-10 -4.50E- 11 -4.50E-11 SI High PLOCA None Low 38 4 -34 1.70E-11 1.70E-11 1.70E-12 1.70E-12 SI High PILOCA-OC None Low 0 2 2 -3.OOE-10 -3.00E-10 -3.00E-11 -3.00E- 11 SI High PILOCA-IC None Low 0 17 17 -8.50E-12 -8.5013-12 -8.50E-13 -8.5013-13 SI Low Class 2 LSS NIA Assume Medium 71 0 -71 7.10E-10 7.1013-10 7.10E-I1 7.10E-11 SI TOTAL -4.15E- 1 -4.15E- I -4.15E-12 -4.15E-12 FW High Class 2 FWU-IC TASCS Medium 5 2 -3 -1.80E-09 9.00E-09 -1.80E-10 9.00E-10 FW High Class 2 FWU-OC None Low 7 4 -3 4.5013-10 4.50E-10 4.50E-1 I 4.50E-1 I FW High Class 2 FWU-IC None Low 3 0 -3 4.5013-10 4.5013-10 4.50E-1 1 4.5013- 1 FW High Class 2 FWI-OC None Low 4 16 12 -1.8013-09 -1.8013-09 -1.8013-10 -1.80E-10 FW TOTAL -2.70E-09 8.10E-09 -2.70E-10 8.10E-10 MS Low Class 2 LSS NIA Assume Medium 16 0 -16 1.6013-10 1.60E-10 1.60E-11 1.6013-I MS TOTAL 1.60E-10 1.60E-10 1.60E-1I 1.60E-1 I CTS Low Class 2 LSS NIA Assume Medium 14 0 -14 1.40E-10 1.4013-10 1.4013- 11 1.4013- 11 CTS TOTAL 1.40E-10 1.40E-10 1.40E-1I 1.40E-I I GRANDTOTAL -6.22E-08 -1.42E-08 -6.22E-09 -1.42E-09 Notes I. Systems are defined on Page 19.

2. Only those ASME Section XI 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.

to AEP:NRC:6055-09 Page 30 Table 3.4-2 Risk Impact Analysis Results for Unit 2 Safety Break Location Failure Potential Inspections CDF Impact LERF Impact systemV)

Significance [ DMs Rank SXI( 2 ) RIS-B Delta w/ POD I w/o POD wPOD w/o POD RC High LOCA TASCS, TT Medium 3 2 -1 -5.40E-09 3.00E-09 -5.40E-10 3.OOE-10 RC High LOCA TT, PWSCC Medium I 1 0 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 RC High LOCA TASCS Medium 2 3 1 -1.26E-08 -3.OOE-09 -1.26E-09 -3.00E-10 RC High LOCA TI" Medium 0 1 1 -5.40E-09 -3.00E-09 -5.40E-10 -3.OOE-10 RC High LOCA PWSCC Medium 5 5 0 0.00E+00 0.00E+00 0.00E+00 0.00E+00 RC High LOCA None Low 63 55 -8 1.20E-09 1.20E-09 1.20E-10 1.20E-10 RC High PLOCA None Low 0 0 0 0.OOE+00 0.00E+00 O.OOE+00 0.00E+00 RC TOTAL -2.22E-08 -1.80E-09 -2.22E-09 -1.80E-10 CS High LOCA TT Medium 0 7 7 -3.78E-08 -2.10E-08 -3.78E-09 -2.10E-09 CS High ILOCA TT Medium 0 0 0 0.00E+00 0.00E+00 0.00E+00 0.00E+00 CS High PLOCA TT Medium 0 0 0 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 CS High LOCA None Low 0 0 0 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 CS High PLOCA None Low 0 0 0 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 CS Low Class 2 LSS NIA Assume Medium 12 0 -12 1.20E-10 1.20E-10 1.20E- 11 1.20E-11 CS TOTAL -3.77E-08 -2.09E-08 -3.77E-09 -2.09E-09 RH High LOCA None Low 1 2 1 -1.50E-10 -1.50E-10 -I.50E- 11 -1.50E-11 RIH High PLOCA None Low 6 4 -2 1.00E-12 1.OOE-12 1.00E-13 1.00E-13 RII High Class 2 SDC-IC None Low 4 0 -4 2.00E-12 2.0013-12 2.0013-13 2.00E-13 RH Low Class 2 LSS NIA Assume Medium 10 0 -10 1.00E-10 1.0013-10 1.0013- 1 1.00E-11 RIl TOTAL - 4.70 E-.70E-1 -4.70E-12 -4.70E-12 to AEP:NRC:6055-09 Page 31 Table 3.4-2 (Cont'd)

Risk Impact Analysis Results for Unit 2

_________ Safety Break Location Failure Potential Inspections CDF Impact LERF Impact Systemt) Significance LSXI(2) DMs Rank RISB Delta w/ POD wlo POD w/POD w/o POT SI High PLOCA IGSCC Medium 12 12 0 0.00E+00 0.00E+00 0.0013+00 0.00E+00 SI High LOCA None Low 3 8 5 -7.50E-10 -7.50E-10 -7.50E-11 -7.50E-1 I SI High PLOCA None Low 38 9 -29 1.45E-1 1 1.45 E-11 1.4513-12 1.45E-12 S$ High PILOCA-OC None Low 0 2 2 -3.00E-10 -3.00E-I0 -3.0013-I 1 -3.00E- 11 SI High PILOCA-IC None Low 0 15 15 -7.50E-12 -7.5013-12 -7.50E-13 -7.5013-13 SI Low Class 2 LSS N/A Assume Medium* 64 0 -64 6.40E-10 6.4013-10 6.4013-11 6.4013-11 SI TOTAL -4.03E-10 -4.03E-10 -4.03E-11 -4.03E-1 I FW High Class 2 FWU-IC TASCS Medium 1 2 1 -9.00E-09 -3.0013-09 -9.0013-10 -3.00E-10 FW High Class 2 FWU-OC None Low 6 2 -4 6.00E-10 6.0013-10 6.00E- 11 6.0013-I FW Htigh Class 2 FWU-IC None Low 5 0 -5 7.5013-10 7.5013-10 7.5013- 1 7.5013- 11 FW High Class 2 FWI-OC None Low 4 16 12 -1.8013-09 -1.8013-09 -1.8013-10 -1.8013-10 FW TOTAL -9.45E-09 -3.45E-09 -9.45E-10 -3.45E-10 MS Low Class 2 LSS N/A Assume Medium 17 0 -17 1.7013-10 1.70E-10 1.7013- 1 1.7013-11 MS TOTAL 1.70E-10 1.70E-10 1.70E- I 1.70E-1 I CTS Low Class 2 LSS N/A Assume Medium 7 0 -7 7.00E-I 1 7.0013- 11 7.00E-12 7.0013-12 CTS TOTAL 7.00E-1 1 7.00E-I I 7.OOE-12 7.00E-12 GRAND -6.95E-08 -2.63E-08 -6.95E-09 -2.63E-09 TOTAL Notes I. Systems are defined on Page 19.

2. Only those ASME Section XI 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.

to AEP:NRC:6055-09 Page 32 Table 5-1 Inspection Location Selection Comparison Between ASME Section XI Code and Code Case N-716 for Unit 1 Sstemn'l Safety Significance Break Location Failure Potential Code Weld Section XI Code Case N-716 Hligh Low DMs j Rank Category Count Vol/Sur [Sur Only RISB Other_2 RC LOCA TASCS, TT Medium B-1 6 3 0 2 RC LOCA TI, PWSCC Medium B-F 1 1 0 1 RC " LOCA TASCS Medium B-1 12 2 2 3 RC " LOCA TT Medium B-J 2 0 0 1 RC " LOCA PWSCC Medium B-F 9 9 0 9 RC B-F 12 12 0 4 V LOCA None Low1 B-J 609 48 11I! 47 RC " PLOCA None Low B-J I11 0 3 0 CS V" LOCA TT Medium B-1 28 0 5 7 CS , ILOCA 'T Medium B-1 2 0 0 0 CS " PLOCA TT Medium B-J 2 0 1 0 CS V' LOCA None Low B-J 16 0 7 0 CS PLOCA None Low B-1 22 0 5 0 CS " Class 2 LSS N/A Assume Medium C-F-I 239 11 3 0 RHIv LOCA None Low B-J 3 1 0 2 RIt PLOCA None Low B-i 19 5 0 3 R V" Class 2 SDC-IC None Low C-F-I 26 4 0 0 RItI " Class 2 LSS N/A Assume Medium C-F-I 282 12 0 0 SI V PLOCA IGSCC Medium B-J 47 13 2 14 SI LOCA None Low B-J 32 5 6 8 SI " PLOCA None Low B-i 232 38 33 4 SI " PILOCA-OC None Low B-J 9 0 2 2 SI V PILOCA-IC None Low B-J 122 0 21 17 SI " Class 2 LSS N/A Assume Medium C-F-I 586 71 0 0 to AEP:NRC:6055-09 Page 33 Table 5-1 (Cont'd)

Inspection Location Selection Comparison Between ASME Section XI Code and Code Case N-716 for Unit 1 Significance Code C- T Section XI r Code Case N-716 t Safety Significance Failure Potential Weld Section XT Code Case N-716 System ') Break Location Code Category Count lligh Low DMs Rank Vol/Sur Sur Only RISB Other(')

FW V.

I Class 22 FWU-IC r

TASCS TASCS 1

Medium T C-F-2 5 8 1 5 1 0 T 2 Class FWU-IC Medium C-F-2 8 5 0 2 FW I/ Class 2 FWU-OC None Low C-F-2 97 7 0 4 FW Class 2 FWU-IC None Low C-F-2 46 3 0 0 FW " Class 2 FWI-OC None Low C-F-2 63 4 0 16 MS " Class 2 LSS N/A Assume Medium C-F-2 218 16 0 0 CTS ' Class 2 LSS NIA Assume Medium C-F-I 165 14 0 0 Notes

1. Systems arc defined on Page 19.

2. The column labeled "Other" is generally used to identify plant augmented inspection program locations credited per Section 4 of Code Case N-716. Code Case N-716 allows the existing plant augmented inspection program for IGSCC (Categories B through G) in a BWR to be credited toward the 10% requirement. This option is not applicable for the CNP RISB application. The "Other" column has been retained in this table solely for uniformity purposes with other RISB application template submittals.

to AEP:NRC:6055-09 Page 34 Table 5-2 Inspection Location Selection Comparison Betwveen ASME Section XI Code and Code Case N-716 for Unit 2 Sstem~') Safety Significance Break Location Failure Potential Code Weld Section XI Code Case N-716 jHigh Low DMs Rank Category Count Vol/Sur Sur Only R1SB IOther(2)

RC V" LOCA TASCS, T7 Medium 3B- 8 3 0 2 RC " LOCA T7, PWSCC Medium B-F 1 1 0 1 RC " LOCA TASCS Medium B-i t0 2 3 3 RC " LOCA TT Medium B-1 2 0 0 1 RC I" LOCA PWSCC Medium B-F 5 5 0 5 B-F 16 16 0 0 RC " LOCA None Low B-i 616 47 115 55 RC " PLOCA None Low B-1 I1 0 3 0 CS " -LOCA TT Medium B-i 29 0 8 7 CS " ILOCA TI1 Medium B-J 3 0 1 0 CS " PLOCA T7 Medium B-i 2 0 0 0 CS I LOCA None Low B-3 14 0 4 0 CS PLOCA None Low B-1 16 0 5 0 CS I Class 2 LSS N/A Assume Medium C-F-I 213 12 5 0 RH " LOCA None Low B-1 3 1 0 2 RI I PLOCA None Low B-i 24 6 0 4 1"

RH Class 2 SDC-IC None Low C-F-I 28 4 0 0 RIH I Class 2 LSS N/A Assume Medium C-F-I 273 10 0 0 i

SI PLOCA IGSCC Medium B-i 43 12 1 12 SI I LOCA None Low B-i 32 3 8 8 SI " PLOCA None Low B-1 239 38 31 9 SI , PILOCA-OC None Low B-i 9 0 4 2 SI " PILOCA-IC None Low B-i3 134 0 21 15 SI " Class 2 LSS N/A Assume Medium C-F-I 545 64 0 0 to AEP:NRC:6055-09 Page 35 Table 5-2 (Cont'd)

Inspection Location Selection Comparison Between ASME Section XI Code and Code Case N-716 for Unit 2 System°' Safety Significance T e 1 Break LocationTIr1T Failure Potential Code Weld Section XI Code Case N-716 High Low DMs Rank Category Count Vol/Sur Sur Only RIS_-B I Other(2 )

FW I Class 2 FWU-IC TASCS Medium C-F-2 8 1 0 2 FW Class 2 FWU-OC None Low C-F-2 93 6 0 2 FW " Class 2 FWU-IC None Low C-F-2 45 5 0 0 FW " Class 2 FWI-OC None Low C-F-2 54 4 0 16 MS " Class 2 LSS NIA Assume Medium C-F-2 211 17 0 0 CTS " Class 2 LSS N/A Assume Medium C-F-I 158 7 0 0 Notes

1. Systems are defined on Page 19.

2. The column labeled "Other" is generally used to identify plant augmented inspection program locations credited per Section 4 of Code Case N-716. Code Case N-716 allows the existing plant augmented inspection program for IGSCC (Categories B through G) in a BWR to be credited toward the 10% requirement. This option is not applicable for the CNP RIS13 application. The "Other" column has been retained in this table solely for uniformity purposes with other RISB application template submittals.

Attachment 2 to AEP:NRC:6055-09 REGULATORY COMMITMENT The following table identifies the action committed to by Indiana Michigan Power Company (I&M) in this document. Any other actions discussed in this submittal represent intended or planned actions by I&M. They are described to the Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments.

Commitment Date I&M will withdraw approved relief requests ISIR-005 and Following approval and ISIR-006. implementation of the risk-informed inservice inspection program.