ML040700258

Risk-Informed Inservice Inspection Program - ASME Code Category B-F, B-J, C-F-1, and C-F-2 Piping

ML040700258
Person / Time
Site:	Farley
Issue date:	03/09/2004
From:	John Nakoski NRC/NRR/DLPM/LPD2
To:	Stinson L Southern Nuclear Operating Co
Peters S, NRR/DLPM, 415-1842
References
TAC MC0178, TAC MC0179
Download: ML040700258 (15)

Text

March 9, 2004 Mr. L. M. Stinson Vice President - Farley Project Southern Nuclear Operating Company, Inc.

Post Office Box 1295 Birmingham, Alabama 35201-1295

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 RE: RISK-INFORMED INSERVICE INSPECTION PROGRAM - ASME CODE CATEGORY B-F, B-J, C-F-1, AND C-F-2 PIPING (TAC NOS. MC0178 AND MC0179)

Dear Mr. Stinson:

By letter dated July 17, 2003, as supplemented by letter dated November 24, 2003, Southern Nuclear Operating Company, Inc., (SNC, the licensee), requested approval of an alternative risk-informed inservice inspection (RI-ISI) program for the Farley Nuclear Plant (FNP), Units 1 and 2 inservice inspection (ISI) program for American Society for Mechanical Engineers (ASME) Class 1 and 2 piping welds.

A proposed RI-ISI program developed in accordance with Westinghouse Owners Group topical report WCAP-14572, Revision 1-NP-A, Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report, is an alternative to the current ASME Section XI ISI program and is applicable to Class 1 and 2 piping at FNP, Units 1 and 2.

The results of the Nuclear Regulatory Commission (NRC) staff review indicate that the licensees proposed RI-ISI program is consistent with WCAP-14572, Revision 1-NP-A and is an acceptable alternative to the requirements of the ASME Code,Section XI for ISI of Code Class 1 and 2 piping, Categories B-F, B-J, C-F-1, and C-F-2 welds.

The NRC staff has reviewed the information provided for this relief request. The NRC staffs Safety Evaluation is provided in the Enclosure. Pursuant to Title 10 of the Code of Federal Regulations, Section 50.55a(a)(3)(i), the NRC grants relief for the second 10-year ISI interval on the basis that the proposed alternative provides an acceptable level of quality and safety.

Sincerely,

/RA/

John A. Nakoski, Chief, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364

Enclosure:

As stated cc w/encl: See next page

Mr. L. M. Stinson March 9, 2004 Vice President - Farley Project Southern Nuclear Operating Company, Inc.

Post Office Box 1295 Birmingham, Alabama 35201-1295

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 RE: RISK-INFORMED INSERVICE INSPECTION PROGRAM - ASME CODE CATEGORY B-F, B-J, C-F-1, AND C-F-2 PIPING (TAC NOS. MC0178 AND MC0179)

Dear Mr. Stinson:

By letter dated July 17, 2003, as supplemented by letter dated November 24, 2003, Southern Nuclear Operating Company, Inc., (SNC, the licensee), requested approval of an alternative risk-informed inservice inspection (RI-ISI) program for the Farley Nuclear Plant (FNP), Units 1 and 2 inservice inspection (ISI) program for American Society for Mechanical Engineers (ASME) Class 1 and 2 piping welds.

A proposed RI-ISI program developed in accordance with Westinghouse Owners Group topical report WCAP-14572, Revision 1-NP-A, Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report, is an alternative to the current ASME Section XI ISI program and is applicable to Class 1 and 2 piping at FNP, Units 1 and 2.

The results of the Nuclear Regulatory Commission (NRC) staff review indicate that the licensees proposed RI-ISI program is consistent with WCAP-14572, Revision 1-NP-A and is an acceptable alternative to the requirements of the ASME Code,Section XI for ISI of Code Class 1 and 2 piping, Categories B-F, B-J, C-F-1, and C-F-2 welds.

The NRC staff has reviewed the information provided for this relief request. The NRC staffs Safety Evaluation is provided in the Enclosure. Pursuant to Title 10 of the Code of Federal Regulations, Section 50.55a(a)(3)(i), the NRC grants relief for the second 10-year ISI interval on the basis that the proposed alternative provides an acceptable level of quality and safety.

Sincerely,

/RA/

John A. Nakoski, Chief, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364

Enclosure:

As stated DISTRIBUTION:

PUBLIC PDII-1 R/F G.Hill (4) OGC cc w/encl: See next page SRosenberg,EDO BBonser,RII SPeters ACRS CHawes GGeorgiev SDinsmore Adams Accession No. ML040700258 **See Previous Concurrence DOCUMENT NAME: G:\PDII-1\Farley\LTR MC0178.wpd *No Major Changes to SE OFFICE PDII-1/PM PDII-1/LA EMCB/SC SPSB/SC OGC PDII-1/SC NAME SPeters** CHawes** TChan* MRubin* MLemoncelli** JNakoski DATE 03/04/04 03/04/04 03/04/04 02/13/04 03/08/04 3/9/04 OFFICIAL RECORD COPY

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO RISK-INFORMED INSERVICE INSPECTION PROGRAM ASME CODE CATEGORY B-F, B-J, C-F-1, AND C-F-2 PIPING SOUTHERN NUCLEAR OPERATING COMPANY, INC., ET AL.

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364

1.0 INTRODUCTION

By letter dated July 17, 2003 (Ref. 1), as supplemented by letter dated November 24, 2003 (Ref. 2), Southern Nuclear Operating Company, Inc., (SNC, the licensee), requested approval of an alternative risk-informed inservice inspection (RI-ISI) program for the Farley Nuclear Plant (Farley or FNP), Units 1 and 2 inservice inspection (ISI) program for American Society of Mechanical Engineers (ASME) Class 1 and 2 piping welds. The current ISI requirements for FNP, Units 1 and 2, are contained in the 1989 Edition of Section XI, Division 1 of the ASME Boiler and Pressure Vessel Code, entitled Rules for Inservice Inspection of Nuclear Power Plant Components (hereinafter called Code). The licensee developed its RI-ISI program in accordance with the Westinghouse Owners Group (WOG) Topical Report WCAP-14572, Revision 1-NP-A (WCAP) (Ref. 3), that the Nuclear Regulatory Commission (NRC) staff has previously reviewed and approved.

2.0 REGULATORY EVALUATION

2.1 Applicable Requirements Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(g) requires that the ISI of ASME Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Code and applicable edition and addenda as required, except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). 10 CFR 50.55a(a)(3) states, in part, that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if the applicant demonstrates that: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, Rules for In-service Inspection (ISI) of Nuclear Power Plant Components, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that in-service examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by

reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein.

2.2 Summary of Proposed Approach In the proposed RI-ISI program, the licensee determined piping failure potential estimates using a software program contained in Supplement 1 to Reference 3, entitled Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed Inservice Inspection, that utilizes probabilistic fracture mechanics technology, industry piping failure history, plant-specific piping failure history, and other relevant information. Using the failure potential and supporting insights on piping failure consequences from the licensees probabilistic risk assessment (PRA), safety significance rankings of piping segments were established to determine inspection locations. The RI-ISI program maintains the fundamental requirements of the Code, such as the examination technique, frequency, and acceptance criteria. However, the RI-ISI program is intended to reduce the number of required examination locations significantly while maintaining an acceptable level of quality and safety.

The licensee plans to implement the RI-ISI program by performing the examinations required under the program during planned outages of the first inspection period (Unit 2), and the third period (Unit 1), of the third 10-year ISI intervals at FNP. The third inspection intervals are scheduled to end on November 30, 2007, and July 29, 2011, for Units 1 and 2, respectively.

Other non-related portions of the Code requirements, as well as the ongoing augmented inspection programs at both units of FNP, will remain unchanged. The RI-ISI program follows a previously approved methodology delineated in Reference 3.

3.0 TECHNICAL EVALUATION

Pursuant to 10 CFR 50.55a(a)(3), the NRC staff has reviewed and evaluated the licensees proposed RI-ISI program, including those portions related to the applicable methodology and processes contained in Reference 3, based on guidance and acceptance criteria provided in Regulatory Guides (RGs) 1.174 (Ref. 4) and 1.178 (Ref. 5) and in the Standard Review Plan (SRP) Chapter 3.9.8 (Ref. 6).

3.1 Proposed Changes to the ISI Program The scope of the licensees proposed RI-ISI program is limited to ASME Class 1 and Class 2 piping only, consisting of Examination Categories B-F and B-J (Class 1) welds, and Categories C-F-1 and C-F-2 (Class 2) welds. SNC proposed the RI-ISI program as an alternative to the existing ISI program that is based on the requirements of the Code. The licensee provided a general description of the proposed changes to the ISI program in Sections 3 and 5 of the submittal (Ref. 1). In revised Tables 5-1a and 5-1b of Reference 2, SNC provided a comparison of inspection location selection between the current ISI program and the proposed RI-ISI program. Similar tables in Reference 1 do not include three additional, Inconel 52 welds that the licensee added to the examination scope in Reference 2. The NRC staff finds that the information submitted adequately defines the proposed changes resulting from the RI-ISI program.

3.2 Engineering Analysis In accordance with the guidance provided in RGs 1.174 and 1.178 (Refs. 4 and 5), the licensee provided the results of an engineering analysis of the proposed changes, using a combination of traditional engineering analysis and PRA. SNC stated that 48 reactor coolant system segments have been identified as high safety-significant (HSS) and RI-ISI examinations on these segments, plus the system pressure test and visual VT-2 examinations, satisfy defense-in-depth considerations. The NRC staff concurs that these inspections are an important element in maintaining the reliability of the reactor coolant system (RCS) as an independent barrier to fission product release. No changes to the evaluation of design basis accidents in the final safety analysis report are being made by the RI-ISI process. Therefore, sufficient safety margins will be maintained.

The licensee stated that the applicable aspects of the ASME Code not affected by the proposed alternative RI-ISI program and the ongoing augmented inspection programs will be retained.

This retention is consistent with the approved WCAP-14572, Revision 1-NP-A; therefore, the NRC staff finds it acceptable.

WCAP-14572, Revision 1-NP-A states, in part, that the SRRA computer models are to be used to estimate the failure probabilities of the structural elements in each of the piping segments. In Reference 1, SNC stated that the failure probabilities for Farley piping segments were all derived using the SRRA software program. The consequence of each segment break is based on the direct and indirect effects of the segment failure. These methods are consistent with the guidelines in WCAP-14572, Revision 1-NP-A, and are in conformance with SRP 3.9.8.

The licensee stated in Reference 1 that an engineering team was established having expertise in the following areas: ISI, non-destructive examination, materials, stress analysis, and system engineering. The engineering team was trained in the failure probability assessment methodology and the Westinghouse SRRA code, including the identification of the software capabilities and limitations as described in WCAP-14572, Revision 1-NP-A. The licensee also stated that the effects of existing augmented ISI programs were included in the risk evaluations and were used in categorizing the segments as described in the approved WCAP-14572, Revision 1-NP-A. When the SRRA code is used to calculate failure probabilities for flow accelerated corrosion (FAC), the licensee used EPRIs CHECKWORKSTM program and plant specific FAC wall-thinning monitoring data to develop SRRA program input.

In Reference 1, SNC stated that FNP, Units 1 and 2 each have a total of 18 dissimilar metal welds located in 15 segments. All the dissimilar metal welds are located in the reactor coolant system piping and are in contact with primary coolant. Six reactor pressure vessel (RPV) nozzle-to-safe end welds and six pressurizer nozzle-to-safe end welds contain Inconel 82 weld material. The remaining six dissimilar metal welds consist of Inconel 52-buttered hot and cold leg nozzles located on the steam generators (SG). Because of primary water stress corrosion cracking (PWSCC) issues associated with Inconel weld material in contact with reactor coolant, the licensee selected all Inconel 82 welds and three of the cold leg SG nozzle Inconel 52 welds for examination. The licensee stated that since Inconel 52 weld material is generally considered to be less susceptible to PWSCC, the remaining three SG Inconel 52 welds, located in the same hot leg as three Inconel 82 RPV outlet nozzle welds, were not selected for examination.

The NRC staff acknowledges that, at present, existing data suggest that Alloy 52 weld material may offer improved resistance to PWSCC over that of Alloy 82 material. However, this understanding is based on a limited amount of data on laboratory prepared specimens.

Further, very little service experience has been accumulated for these weld materials in thick section reactor coolant piping for pressurized water reactors. Recent investigations found that many weldability issues associated with Alloy 52/152 thick welds are just beginning to be recognized. Significant amounts of ductility dip cracking, lack of fusion and porosity have been observed. Weldability issues like these have resulted in significant numbers of repairs and higher localized residual stresses at the inside surface of the weld. In NRC special report 50-395/00-08, dated March 15, 2001, an NRC inspection team concluded that the PWSCC phenomenon for Alloy 52/152 welding material is not fully understood and further studies developing quantitative data should be performed before the new Alloy 52/152 weld can be considered immune to PWSCC. In light of these facts, and in keeping with fundamental defense-in-depth principles, the NRC staff considers that PWSCC should be treated as a potential degradation mechanism for all Alloy 52/152 and 82/182 dissimilar metal welds in contact with primary system coolant. In Reference 2, the licensee agreed to include the three additional Inconel 52 welds into the 10-year RI-ISI examination scope. The licensee also indicated that a re-evaluation of any new quantitative data concerning the resistance of heavy wall Inconel 52 welds to PWSCC will be performed to determine if the three Inconel 52 welds should continue to be in the examination scope during the next 10-year interval RI-ISI program update. Because the licensee is treating PWSCC as a potential degradation mechanism for Alloy 52/152 and 82/182 dissimilar metal welds, the NRC staff finds this approach acceptable.

The licensee reported two deviations in Reference 1 and one deviation in Reference 2 from the WCAP-14572, Revision 1-NP-A, methodology. The deviations are 1) credit taken for leak detection when calculating pipe failure probabilities, 2) the evaluation of the potential impact of parameter uncertainty, and 3) performing visual VT-2 examinations in lieu of volumetric examinations for HSS socket welds and branch connections 2 inches nominal pipe size (NPS) and smaller. The licensee also identified a discrepancy in Table 4.1-1 of the WCAP that, for welds subject to PWSCC, only requires a visual VT-2 examination. Footnote 7 further defines the volume of material that should be examined. Visual VT-2 examinations are not volumetric examinations, therefore, SNC will perform volumetric examinations for welds subject to PWSCC, and when specified elsewhere in Table 4.1-1, visual VT-2 examinations will be performed in accordance with Table IWB-2500-1 of the ASME Code. The NRC staff recognizes the error in Table 4.1-1 as related to PWSCC and visual VT-2 examinations, and since SNC proposed to perform the proper volumetric examinations for all dissimilar metal welds at FNP, Units 1 and 2, the NRC staff concludes that the licensees approach is acceptable.

WCAP-14572, Revision 1-NP-A allows credit for detecting (and isolating, repairing, or otherwise terminating a potential accident sequence) a leak in the RCS piping before it develops into a pipe break for piping inside the containment. This credit reflects the highly developed leak detection systems used to monitor leakage from the reactor coolant piping. Detection of a leak before break is plausible for any non-RCS segment located inside the containment that interfaces with the RCS by use of radiation and sump level monitors that can detect a leak in the segment as reliably as that of an RCS leak. The licensee identified non-RCS segments inside the containment and credited leak detection for these segments. Since the segments are subject to essentially the same leak detection capabilities as that of an RCS leak, the extension of credit for leak detection in these segments is reasonable and acceptable.

The WCAP states that an initial calculation of the risk reduction worth (RRW) using point estimate input values should be followed by a sensitivity study that assigns uncertainty to the input values. The aim of the sensitivity analysis is to investigate the potential movement of segments from low to high safety significance based on the uncertainty of quantitative inputs and the guideline values defining the low, medium and high RRW ranges. Instead of performing a uncertainty analysis as a sensitivity study, the licensee incorporated the uncertainty analysis directly into the initial calculation of the RRW values. This process also identifies any segments that might move to higher safety-significance based on the uncertainty in the inputs. The NRC staff finds the licensees approach to uncertainties to be acceptable.

The last deviation regards the licensees request to perform VT-2 inspections and surface examinations (as appropriate) in lieu of volumetric examinations of HSS socket welds 2 inches NPS and smaller. The licensee provided the basis for this deviation in Reference 2. The WCAP-14572, Revision 1-NP-A, Table 4.1-1, Category R-A, Item R1.11, requires a volumetric examination for elements (i.e., welds) selected for examination that are subject to thermal fatigue. This requirement encompasses the examination of socket welds as well as butt welds.

Because volumetric examination of partial penetration socket welds for detection of cracks originating at the weld root is impractical due to the geometric configuration of the welded joint, the licensee proposed to perform VT-2 visual examinations and surface examinations, as applicable, for Item R1.11 HSS socket welds. A list of the affected HSS socket-welded segments identified in Reference 2 is shown in Table 3.2-1 below.

Table 3.2-1 Farley Risk-Informed ISI HSS Socket-Welded Segments for Proposed Deviation Segments NPS Description RI-ISI Comments Exam RC-012A, B 3/8-3/4 Instrument Lines VT-2 Not subject to OD cracking. The RI-ISI exam is RC-020A, B, C 3/8-3/4 the same as the Section XI required VT-2 exam.

RC-054 1 Thermowell VT-2 Not subject to OD cracking. The RI-ISI exam is the same as the Section XI required VT-2 exam.

RC-017 1 Capped Lines VT-2 Not subject to OD cracking. The RI-ISI exam is the same as the Section XI required VT-2 exam.

RC-018A, B, C 1 RTDs VT-2 Not subject to OD cracking. The RI-ISI exam is the same as the Section XI required VT-2 exam.

RC-032A, B, C 3/4 Vents, Drains, and VT-2 Not subject to OD cracking. The RI-ISI exam is RC-036A, B, C Test Connections the same as the Section XI required VT-2 exam.

RC-041A RC-024B, C 2 RC Drain to Drain Surface Potentially subject to OD cracking. The RI-ISI Tank exam is the same as the Section XI required surface exam.

RC029A, B, C 2 Safety Injection Surface Potentially subject to OD cracking. The RI-ISI Lines exam is the same as the Section XI required surface exam.

RC-059 1 Excess Letdown Surface Potentially subject to OD cracking. The RI-ISI Line exam is the same as the Section XI required surface exam.

The licensee pointed out that this deviation is consistent with proposed ASME Section XI Code, "Appendix X, Risk-Informed Requirements for Piping," for Item R1.11 socket-welded segments and branch piping connection welds (NPS 2 and smaller). Appendix X replaces the volumetric examination with a visual VT-2 examination, while the proposed deviation replaces the volumetric examination with a visual VT-2 examination plus surface examinations, as

appropriate. The NRC staff notes that the proposed Appendix X incorporates the RI-ISI examination requirements previously specified in ASME Code Case 577-1. While neither Code Case N-577-1 nor proposed Appendix X have been reviewed or approved by the NRC, the NRC staff acknowledges that geometric limitations associated with socket welds may mask small cracks initiating at the root of the partial penetration weld and in many cases render volumetric examinations ineffective. Therefore, for thermal fatigue degradation mechanisms that induce cracking at the root of socket-welds in HSS 2-inch NPS or less, visual VT-2 examinations are reasonable.

The licensee also noted that the existing ASME Section XI Code requirements list surface examination, not volumetric examination, for socket welds. SNC stated, and the NRC staff agrees, that while many socket weld failures result from cracking initiated at the weld root, surface examinations can be an effective method for the identification of outside surface initiated cracking, specifically flaws induced by low-cycle thermal fatigue or external chloride stress corrosion cracking (ECSCC). The licensee evaluated the potential for these conditions in each HSS socket-welded segment in Table 3.2-1. SNC indicated that material controls and cleanliness requirements at FNP have prevented and should continue to prevent outside diameter (OD) initiated ECSCC on these socket-welded segments. SNC concluded that piping configurations in vents, drains, test connections, capped lines, instrument lines, and RTD-type segments preclude the occurrence of thermal fatigue mechanisms that would result in low-cycle bending stresses outside design tolerances. However, the licensee indicated that a limited potential for low-cycle, high bending stress thermal fatigue exists in excess letdown segment RC-059, safety injection segments RC-029 A, B, and C, and reactor coolant drain to drain tank segments RC-024 B and C. Because of these segments potential for OD-initiated cracking, the licensee indicated that it will perform a surface examination on one weld in each of these socket-welded segments.

The NRC staff notes that low-cycle, high bending stress, thermal fatigue is primarily associated with large global bending loading due to the presence of cyclic thermal stratification conditions.

The potential for these conditions is dependent upon pipe routing and support configurations, pipe sizes, and operating conditions. Generally, this condition would not be expected to occur in piping less than 2-inch NPS with the configurations described by the licensee. Therefore, the conclusions reached by the licensee regarding the potential for OD-initiated cracking in the affected socket-welded segments are reasonable. The NRC staff agrees that, for OD-initiated cracking, a surface examination is appropriate and since the licensees evaluation identified only a potential (no active mechanism) for this condition, performing a surface examination on at least one weld in each of the affected HSS segments (in addition to the VT-2 examinations performed each refueling outage) is consistent with the recommendations in WCAP-14572, Revision 1-NP-A.

Based on the above evaluation, the NRC staff believes that the proposed deviation to perform VT-2 visual inspections and limited surface examinations (as appropriate) in lieu of the volumetric examination requirements described in WCAP-14572, Revision 1-NP-A, Table 4.1-1, Category R-A, Item R1.11, for HSS socket-welded segments 2-inch NPS and smaller is acceptable and will provide reasonable assurance of structural integrity for these segments.

3.3 Probabilistic Risk Assessment The licensee used Revision 5 of the Farley Level 1 and Level 2 PRA models to evaluate the impacts of this change on plant risk. The estimated core damage frequency (CDF) and large early release frequency (LERF) for Unit 1 are 3.86E-5/year and 4.19E-7/year, respectively, whereas the estimated CDF and LERF for Unit 2 are 5.81E-5/year and 4.26E-7/year, respectively. The individual plant examination (IPE) model was submitted in June, 1993, and the NRC staff evaluation report for the IPE was issued on February 26, 1996. In August 2001, WOG conducted a peer review on Revision 4 of the PRA.

The WOG and NRC staff reviewers provided about 25 observations and comments on the specific methods or application of the methods used by the licensee in the IPE and Revision 4 of the PRA. In Reference 1 and Reference 2, the licensee discusses the observations developed by the WOG PRA peer reviewers and the NRC staff respectively. The licensee modified the PRA in response to nine of the observations. SNC further explained why several of the observations are based on differences between equally valid assumptions and methods and do not require modifications to the PRA. The licensee also explained that the possible PRA modifications to resolve the remaining observations would have an insignificant impact on the RI-ISI program. After reviewing the licensees discussion, the NRC staff concurs that additional modifications to resolve the issues would have an insignificant impact on the proposed RI-ISI program.

The NRC staff did not review the PRA analysis to assess the accuracy of the quantitative estimates. Quantitative results of the PRA are used, in combination with a quantitative characterization of the pipe segment failure likelihood, to support the assignment of segments into broad safety significance categories reflecting the relative importance of pipe segment failures on CDF and LERF. Inaccuracies in the models or assumptions large enough to invalidate the broad categorizations developed to support the RI-ISI should have been identified in the licensees or in the NRC staffs review. Minor errors or inappropriate assumptions will only affect the consequence categorization of a few segments and will not invalidate the general results or conclusions. The continuous use and maintenance of the PRA provides further opportunities to identify inaccuracies and inappropriate assumptions, if any, in the PRA models. The NRC staff finds that the quality of the PRA is sufficient to support the submittal.

The reported changes in CDF and LERF are provided in the following Table 3.3.

Table 3.3 - Changes in CDF and LERF as a Function of Operator Action Unit 1 Unit 2 CDF without operator action -6E-7/year -5E-7year CDF with operator action -7E-8/year -3E-8/year LERF without operator action -3E-9/year -2E-9year LERF with operator action -2E-10/year -1E-10/year

The licensee did not submit estimates for the other risk change criteria in Section 4.4.2 of WCAP-14572, Revision 1-NP-A, but stated in Reference 1 that all the change in risk calculations were performed according to the guidance on page 213 of the WCAP-14572, Revision 1-NP-A (as applicable), and all four criteria for evaluating the results were applied.

Based on the use of the approved methodology and on the reported results, the NRC staff finds that any change in risk associated with the implementation of the RI-ISI program is small and consistent with the intent of the Commissions Policy Statement (Ref. 7) and, therefore, is consistent with RG 1.178.

3.4 Integrated Decisionmaking The proposed RI-ISI program presents an integrated approach that considers, in concert, the traditional engineering analysis, the risk evaluation, and the implementation and performance monitoring of piping. The selection of pipe segments to be inspected is described in Section 3.8 of Reference 1 using the results of the risk category rankings and other operational considerations. Revised Tables 5-1a and 5-1b of Reference 2 provide a summary comparing the number of inspections required under the existing ASME Section XI ISI program at FNP, Units 1 and 2 with the alternative RI-ISI program. The WCAP-14572, Revision 1-NP-A, methodology includes a statistical calculation that is applied to determine the number of examinations required in the population of HSS welds, excluding susceptible locations, to satisfy certain statistical criteria. This approach is consistent with the concept that, by focusing inspections on the most safety significant welds, the number of inspections can be reduced while at the same time maintaining public health and safety. Therefore, this approach is acceptable. The NRC staff finds that the licensees selection process uses defense-in-depth considerations and is consistent with the WCAP-14572, Revision 1-NP-A.

The objective of ISI required by the Code is to identify service-induced conditions (i.e., flaws or other degradation) that may challenge the structural integrity of components and adversely impact plant safety. Therefore, the RI-ISI program must meet this objective to be found acceptable for use. Further, since the RI-ISI program is partially based on an inspection for cause philosophy, examination element selection should target specific degradation mechanisms.

Section 4 of WCAP-14572, Revision 1-NP-A, provides guidelines for the areas and/or volumes to be inspected as well as the examination method, acceptance standard, and evaluation criteria for each degradation mechanism. Based on a review of the cited portion of WCAP-14572, Revision 1-NP-A, the NRC staff concludes that the examination methods are appropriate since they are selected based on specific degradation mechanisms, pipe sizes, and materials of concern. The licensee stated in Reference 2 that, unless NRC-approved relief has been granted, all requirements in Table 4.1-1 of WCAP-14572, Revision 1-NP-A, will be implemented. The licensee requested one deviation to Table 4.1-1 that will be applied to socket welds less than 2-inch NPS. The Table requires volumetric examination for elements subject to thermal fatigue, however, the licensee will perform visual VT-2 examinations in lieu of the specified volumetric examinations for these small-bore piping welds. The NRC staff evaluated this deviation in Section 3.2 above.

3.5 Implementation and Monitoring Implementation and performance monitoring strategies require careful consideration by the licensee and are addressed in Element 3 of RG 1.178 and SRP 3.9.8. The objective of Element 3 is to assess performance of the affected piping systems under the proposed RI-ISI program by implementing monitoring strategies that confirm the assumptions and analyses used in the development of the RI-ISI program. To approve an alternative pursuant to 10 CFR 50.55a(a)(3)(i), implementation of the RI-ISI program, including inspection scope, examination methods, and methods of evaluation of examination results, must provide an adequate level of quality and safety.

In Reference 1, the licensee stated that upon approval of the RI-ISI program, procedures that comply with the WCAP-14572, Revision 1-NP-A, guidelines will be prepared to implement and monitor the RI-ISI program. The licensee confirmed that the applicable portions of the Code not affected by the change, e.g., inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements, would be retained.

SNC stated in Section 4 of Reference 1 that the RI-ISI program is a living program and its implementation will require feedback of new relevant information to ensure the appropriate identification of HSS piping locations. Reference 1 also stated that, as a minimum, risk ranking of piping segments will be reviewed and evaluated every ISI period and that significant changes may require more frequent adjustments as directed by any NRC Bulletin or Generic Letter or by industry and plant-specific feedback. The NRC staff finds that the proposed process for RI-ISI program updates meets the guidelines of RG 1.174 that risk-informed applications should include performance monitoring and feedback provisions; therefore, the process for program updates and monitoring is acceptable.

The licensees existing third 10-year ISI programs are, at present, within the first 40-month inspection period at Unit 2, and the third 40-month inspection period at Unit 1. The licensee intends to integrate the RI-ISI program into the existing Code ISI programs, i.e., the RI-ISI program will supercede the Code for selection of piping welds in Categories B-F, B-J, C-F-1 and C-F-2. In Reference 1 the licensee provides an implementation schedule that includes one-third of the RI-ISI examinations to be completed by the end of the current inspection periods at each Unit. The completion of approximately one-third of all examinations during each of the inspection periods is the same as existing Code requirements. To determine the selection and distribution of locations for these examinations, variables such as failure mechanisms, industry and site-specific experience, inspection history, and stress were considered. The NRC staff finds the logic used for selecting the extent of examinations to be performed during the current periods of the third 10-year intervals at FNP, Units 1 and 2 to be consistent with ASME requirements and is, therefore, acceptable.

4.0 CONCLUSION

10 CFR 50.55a(a)(3)(i) permits alternatives to regulatory requirements when authorized by the NRC if the applicant demonstrates that the alternative provides an acceptable level of quality and safety. In this case, the licensees proposed alternative is to use the RI-ISI process described in the NRC-approved WCAP-14572, Revision 1-NP-A. The licensee identified three deviations from the approved methodology: inclusion of the parameter uncertainty in the initial

calculation of the RRW values, crediting of leak detection for several non-RCS piping segments, and performing VT-2 visual and limited surface examinations in lieu of volumetric examinations of socket welds 2 inches NPS and smaller. The NRC staff has determined that these deviations are acceptable as discussed in Section 3.2 of this Safety Evaluation.

The NRC staff finds that the results of different elements of the engineering analysis are considered in an integrated decision-making process. The impact of the proposed changes in the ISI program is founded on the adequacy of the engineering analysis and acceptable estimation of changes in plant risk in accordance with RG 1.174 and RG 1.178 guidelines.

The SNC methodology also considers implementation and performance monitoring strategies.

Inspection strategies ensure that failure mechanisms of concern have been addressed and there is adequate assurance of detecting damage before structural integrity is affected. The risk significance of piping segments is taken into account in defining the inspection scope for the RI-ISI program.

System pressure tests and visual examination of piping structural elements will continue to be performed on all Code Class 1 and 2 systems in accordance with the ASME Code Section XI program. The RI-ISI program applies the same performance measurement strategies as existing ASME Code requirements.

The SNC risk-informed methodology provides for conducting an analysis of the proposed changes using a combination of engineering analysis with supporting insights from a PRA.

Defense-in-depth and quality are not degraded in that the methodology provides reasonable confidence that any reduction in existing inspections will not lead to degraded piping performance when compared to existing performance levels. Inspections are focused on locations with active degradation mechanisms as well as selected locations that monitor the performance of system piping.

As discussed above, the NRC staffs review of the licensees proposed RI-ISI program concludes that it is an acceptable alternative to the current ISI program for Code Class 1, Categories B-F and B-J piping welds, and for Code Class 2, Categories C-F-1 and C-F-2 piping welds. In addition, the licensee has met the applicable criteria described in SRP 3.9.3. Based on risk considerations and the criteria of the SRP, the NRC staff concludes that the licensee's proposed alternative will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the licensees proposed RI-ISI program is authorized for the remainder of the third 10-year inspection intervals at FNP, Units 1 and 2. All other ASME Code Section XI requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

5.0 REFERENCES

1. Letter from J. B. Beasley, Jr., Southern Nuclear Operating Company, Inc., to USNRC, Joseph M. Farley Nuclear Plant, Risk-Informed Inservice Inspection Program, ASME Code Category B-F, B-J, C-F-1, and C-F-2 Piping, July 17, 2003.

2. Letter from J. B. Beasley, Jr., Southern Nuclear Operating Company, Inc. to USNRC, Joseph M. Farley Nuclear Plant Risk Informed Inservice Inspection Program Submittal, Response to a Request for Additional Information, November 24, 2003.

3. WCAP-14572, Revision 1-NP-A, Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report, February 1999.

4. NRC Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, July 1998.

5. NRC Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decision Making: Inservice Inspection of Piping, September 1998.

6. NRC NUREG-0800, Chapter 3.9.8, Standard Review Plan for Trial Use for the Review of Risk-Informed Inservice Inspection of Piping, May 1998.

7. USNRC, Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities; Final Policy Statement, Federal Register, Vol. 60, p. 42622, August 16, 1995.

Principal Contributors: G. Georgiev S. Dinsmore Date: March 9, 2004

Joseph M. Farley Nuclear Plant cc:

Mr. Don E. Grissette William D. Oldfield General Manager - SAER Supervisor Southern Nuclear Operating Company Southern Nuclear Operating Company Post Office Box 470 P. O. Box 470 Ashford, Alabama 36312 Ashford, Alabama 36312 Mr. B. D. McKinney, Licensing Manager Southern Nuclear Operating Company Post Office Box 1295 Birmingham, Alabama 35201-1295 Mr. M. Stanford Blanton Balch and Bingham Law Firm Post Office Box 306 1710 Sixth Avenue North Birmingham, Alabama 35201 Mr. J. B. Beasley, Jr.

Executive Vice President Southern Nuclear Operating Company Post Office Box 1295 Birmingham, Alabama 35201 State Health Officer Alabama Department of Public Health 434 Monroe Street Montgomery, Alabama 36130-1701 Chairman Houston County Commission Post Office Box 6406 Dothan, Alabama 36302 Resident Inspector U.S. Nuclear Regulatory Commission 7388 N. State Highway 95 Columbia, Alabama 36319