ML18086B110

Relief Requests Associated with the Use of Code Cases N-513-4 and N-513-3 for the Fifth Inservice Inspection Interval

ML18086B110
Person / Time
Site:	Peach Bottom
Issue date:	03/26/2018
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RR 15R-07, RR 15R-08
Download: ML18086B110 (14)

Text

Exelon Generation © 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.55a March 26, 2018 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 N RC Docket Nos. 50-277 and 50-278

Subject:

Relief Requests Associated with the Use of Code Cases N-513-4 and N-513-3 for the Fifth lnservice Inspection Interval Attached for your review are two relief requests associated with the fifth lnservice Inspection (ISi) interval for the Peach Bottom Atomic Power Station, Units 2 and 3. The fifth interval program complies with the 2013 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code. The fifth ISi interval is scheduled to begin on January 1, 2019 and is currently scheduled to end December 31, 2028. We request your approval of this package by December 31, 2018.

Both relief requests are associated with the use of Code Case N-513 and are being submitted together for use as part of the fifth interval.

There are no regulatory commitments in this letter.

If you have any questions concerning this letter, please contact Tom Loomis at (61 O) 765-5510.

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James Barstow Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC

Attachment:

Relief Requests 15R-07 and 15R-08 cc: USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, PBAPS USNRC Project Manager, PBAPS R. R. Janati, Pennsylvania Bureau of Radiation Protection S. T. Gray, State of Maryland

Attachment Relief Requests ISR-07 and ISR-08

10CFRS0.55a Relief Request ISR-07 (Page 1of5)

Request for Relief to Utilize Code Case N-513-4 in Accordance with 10 CFR 50.55a(z)(2)

1. ASME Code Component(s) Affected All American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV),Section XI, Class 2 and 3 components that meet the operational and configuration limitations of ASME Code Case N-513-4 (N-513-4), "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping,Section XI, Division 1,"

paragraphs l(a), l(b), l(c), and l(d).

2. Applicable Code Edition The fifth 10-year interval of the Peach Bottom Atomic Power Station, Units 2 and 3 Inservice Inspection (ISi) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, 2013 Edition.

3. Applicable Code Requirement IWC-3120 and IWD-3120 of ASME Section XI, require that flaws exceeding the defined acceptance criteria be corrected by repair/replacement activities or evaluated and accepted by analytical evaluation. IWC-3130 and IWD-3130 of ASME Section XI, require that relevant conditions be subject to supplemental examination, corrective measures or repair/replacement activities, or evaluated and accepted by analytical evaluation.

4. Reason for Request In accordance with 10CFR50.55a(z)(2), relief is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Exelon Generation Company, LLC (EGC) is requesting a proposed alternative from the requirement to perform repair/replacement activities for degraded Class 2 and 3 piping whose maximum operating temperature does not exceed 200°F and whose maximum operating pressure does not exceed 275 psig for Peach Bottom Atomic Power Station, Units 2 and 3. Moderately degraded Class 2 and 3 piping could require a plant shutdown within the required action statement timeframes to repair observed degradation. Plant shutdown activities result in additional dose and plant risk that would be inappropriate when a degraded condition is demonstrated to retain adequate margin to complete the component's function. The use of an acceptable alternative analysis method in lieu of immediate action for a degraded condition will allow EGC to perform additional extent of condition examinations on the affected systems while allowing time for safe and orderly long term repair actions if necessary. Actions to remove degraded piping from service could have a detrimental overall risk impact by requiring a plant shutdown, thus

10CFR50.55a Relief Request ISR-07 (Page 2 of 5) requiring use of a system that is in standby during normal operation. Accordingly, compliance with the current Code requirements results in a hardship without a compensating increase in the level of quality and safety.

ASME Code Case N-513-3 (N-513-3) does not allow evaluation of flaws located away from attaching circumferential piping welds that are in elbows, bent pipe, reducers, expanders, and branch tees. N-513-3 also does not allow evaluation of flaws located in heat exchanger external tubing or piping. N-513-4 provides guidance for evaluation of flaws in these locations.

5. Proposed Alternative and Basis for Use EGC is requesting approval to apply the evaluation methods of N-513-4 to Class 2 and 3 components that meet the operational and configuration limitations of N-513-4, paragraphs l(a), l(b), l(c), and l(d) for Peach Bottom Atomic Power Station, Units 2 and 3 in order to avoid accruing additional personnel radiation exposure and increased plant risk associated with a plant shutdown to comply with the cited Code requirements.

The Nuclear Regulatory Commission (NRC) issued Generic Letter 90-05 (Reference 1),

"Guidance for Performing Temporary Non-Code Repair of Class 1, 2, and 3 Piping (Generic Letter 90-05)," to address the acceptability of limited degradation in moderate energy piping. The generic letter defines conditions that would be acceptable to utilize temporary non-Code repairs with NRC approval. The ASME recognized that relatively small flaws could remain in service without risk to the structural integrity of a piping system and developed ASME Code Case N-513 (N-513). NRC approval of N-513 versions in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 18 (Reference 4), allows acceptance of partial through-wall or through-wall leaks for an operating cycle provided all conditions of the code case and NRC conditions are met. The code case also requires the Owner to demonstrate system operability due to leakage.

The ASME recognized that the limitations in N-513-3 were preventing needed use in piping components such as elbows, bent pipe, reducers, expanders, and branch tees and external tubing or piping attached to heat exchangers. N-513-4 was approved by the ASME to expand use on these locations and to revise several other areas of the code case.

Attachment 2 of the Reference 2 letter provides a marked-up N-513-3 version of the code case to highlight the changes compared to the NRC approved N-513-3 version.

Attachment 3 of the Reference 2 letter provides the ASME approved N-513-4. The following provides a high level overview of the N-513-4 changes:

I) Revised the maximum allowed time of use from no longer than 26 months to the next scheduled refueling outage.

2) Added applicability to piping elbows, bent pipe, reducers, expanders, and branch tees where the flaw is located more than (Rot) 112 from the centerline of the attaching circumferential piping weld.

10CFRSO.SSa Relief Request ISR-07 (Page 3 of 5)

3) Expanded use to external tubing or piping attached to heat exchangers.

4) Revised to limit the use to liquid systems.

5) Revised to clarify treatment of Service Level load combinations.

6) Revised to address treatment of flaws in austenitic pipe flux welds.

7) Revised to require minimum wall thickness acceptance criteria to consider longitudinal stress in addition to hoop stress.

8) Other minor editorial changes to improve the clarity of the code case.

Detailed discussion of significant changes in N-513-4 when compared to NRC approved N-513-3 is provided in Attachment 4 of the Reference 2 letter.

The design basis is considered for each leak and evaluated using the EGC Operability Evaluation process. The evaluation process must consider requirements or commitments established for the system, continued degradation and potential consequences, operating experience, and engineering judgement. As required by the code case, the evaluation process considers but is not limited to system make-up capacity, containment integrity with the leak not isolated, effects on adjacent equipment, and the potential for room flooding.

Leakage rate is not typically a good indicator of overall structural stability in moderate energy systems, where the allowable through-wall flaw sizes are often on the order of inches. The periodic inspection interval defined using paragraph 2(e) of N-513-4 provides evidence that a leaking flaw continues to meet the flaw acceptance criteria and that the flaw growth rate is such that the flaw will not grow to an unacceptable size.

The effects of leakage may impact the operability determination or the plant flooding analyses specified in paragraph l(f). For a leaking flaw, the allowable leakage rate will be determined by dividing the critical leakage rate by a safety factor of four (4). The critical leakage rate is determined as the lowest leakage rate that can be tolerated and may be based on the allowable loss of inventory or the maximum leakage than can be tolerated relative to room flooding, among others. The safety factor of four (4) on leakage is based upon ASME Code Case N-705 (N-705) (Reference 3), which is accepted without condition in Regulatory Guide 1.147, Revision 18. Paragraph 2.2(e) of N-705 requires a safety factor of two (2) on flaw size when estimating the flaw size from the leakage rate. This corresponds to a safety factor of four (4) on leakage for non planar flaws. Although the use of a safety factor for determination of an unknown flaw is considered conservative when the actual flaw size is known, this approach is deemed acceptable based upon the precedent of N-705. Note that the alternative herein does not propose to use any portion of N-705 and that citation of N-705 is intended only to provide technical basis for the safety factor on leakage.

10CFR50.55a Relief Request ISR-07 (Page 4 of 5)

During the temporary acceptance period, leaking flaws will be monitored daily as required by paragraph 2(f) of N-513-4 to confirm the analysis conditions used in the evaluation remain valid. Significant change in the leakage rate is reason to question that the analysis conditions remain valid, and would require re-inspection per paragraph 2(f) of the code case. Any re-inspection must be performed in accordance with paragraph 2(a) of the code case.

The leakage limit provides quantitative measurable limits which ensure the operability of the system and early identification of issues that could erode defense-in-depth and lead to adverse consequences.

In summary, EGC will apply N-513-4 to the evaluation of Class 2 and 3 components that are within the scope of the code case. N-513-4 utilizes technical evaluation approaches that are based on principals that are accepted in other Code documents already acceptable to the NRC. The application of this code case, in conjunction with safety factors on leakage limits, will maintain acceptable structural and leakage integrity while minimizing plant risk and personnel exposure by minimizing the number of plant transients that could be incurred if degradation is required to be repaired based on ASME Section XI acceptance criteria only.

6. Duration of Proposed Alternative The proposed alternative is for use of N-513-4 for Class 2 and Class 3 components within the scope of the code case. An ASME Section XI compliant repair/replacement will be completed prior to exceeding the next refueling outage or allowable flaw size, whichever comes first. Relief is requested for the Fifth ISi Interval for Peach Bottom Atomic Power Station, Units 2 and 3, or until the NRC approves N-513-4, or a later revision, in Regulatory Guide 1.147 or other document during the interval. If a flaw is evaluated near the end of the interval for Peach Bottom Atomic Power Station, Units 2 and 3 and the next refueling outage is in the subsequent interval, the flaw may remain in service under this relief request until the next refueling outage.

7. Precedent Peach Bottom Atomic Power Station, Units 2 and 3, fourth ISi interval relief request was authorized by NRC Safety Evaluation (SE) dated September 6, 2016 (Reference 5). This Peach Bottom Atomic Power Station, Units 2 and 3, relief request was part of an EGC fleet-wide submittal, and the alternative for the use of N-513-4 was authorized for various stations. Relief Request 15R-07 for the Peach Bottom Atomic Power Station, Units 2 and 3, fifth ISi interval, utilizes a similar approach to the previously approved relief request.

8. References l) NRC Generic Letter 90-05, "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping (Generic Letter 90-05)," dated June 15, 1990.

10CFRSO.SSa Relief Request ISR-07 (Page 5 of 5)

2) Letter from D. T. Gudger (Exelon Generation Company, LLC) to NRC, "Proposed Alternative to Utilize Code Case N-513-4, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division l,"

dated January 28, 2016.

3) ASME Boiler and Pressure Vessel Code, Code Case N-705, "Evaluation Criteria for Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3 Vessels and TanksSection XI, Division l," dated October 12, 2006.

4) NRC Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division l," Revision 18.

5) Letter from G. E. Miller (NRC) to B. C. Hanson (Exelon Generation Company, LLC), Proposed Alternative to Use ASME Code Case N-513-4 (CAC Nos. MF7301-MF7322), dated September 6, 2016 (ADAMS Accession No. ML16230A237).

10CFR50.55a Relief Request ISR-08 (Page l of 7)

Request for Relief to Utilize Code Case N-513-3 in Accordance with 10 CFR 50.55a(z)(2)

1. ASME Code Component(s) Affected All American Society of Mechanical Engineers (ASME), Boiling and Pressure Vessel (BPV),Section XI, Class 3 High Pressure Service Water (HPSW) System piping that operates at a pressure less than or equal to 375 psig but greater than 275 psig in Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3.

This relief request affects components outside the limitations covered under Relief Request I5R-07.

2. Applicable Code Edition and Addenda The fifth 10-year interval of the Peach Bottom Atomic Power Station, Units 2 and 3 Inservice Inspection (ISI) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, 2013 Edition.

3. Applicable Code Requirement IWD-3120 of ASME Section XI requires that flaws exceeding the defined acceptance criteria be corrected by repair/replacement activities or evaluated and accepted by analytical evaluation. IWD-3130 of ASME Section XI, requires that relevant conditions be subject to supplemental examination, repair/replacement activities, or accepted by analytical evaluation.

4. Reason for Request In accordance with 10CFR50.55a(z)(2), relief is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Exelon Generation Company, LLC (EGC) is requesting a proposed alternative from the requirement to perform repair/replacement activities for degraded HPSW System piping that has a maximum operating pressure in excess of 275 psig. Moderately degraded piping could require a plant shutdown within the required action statement timeframes to repair observed degradation. Plant shutdown activities result in additional dose and plant risk that would be inappropriate when a degraded condition is demonstrated to retain adequate margin to complete the component's function. The use of an acceptable alternative analysis method in lieu of immediate action for a degraded condition will allow EGC to perform additional extent of condition examinations on the affected systems while allowing time for safe and orderly long term repair actions if necessary.

Actions to remove degraded piping from service could have a detrimental overall risk

10CFR50.55a Relief Request ISR-08 (Page 2 of 7) impact by requiring a plant shutdown, thus requiring use of a system that is in standby during normal operation. Accordingly, compliance with the current Code requirements results in a hardship without a compensating increase in the level of quality and safety.

S. Proposed Alternative and Basis for Use EGC is requesting approval to apply the evaluation methods of ASME Code Case N-513-3 (N-513-3), "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1," (Reference 1) to the HPSW System piping having a maximum operating pressure of 375 psig, in order to avoid accruing additional personnel radiation exposure and increased plant risk associated with a plant shutdown to comply with the cited ASME Section XI requirements. The relief request will be applied to HPSW piping with corrosion degradation only if ASME code repairs cannot be reasonably completed within the Technical Specification required time limit.

The Nuclear Regulatory Commission (NRC) issued Generic Letter 90-05, "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping (Generic Letter 90-05)," (Reference 2) to address the acceptability of limited degradation in moderate energy piping. The generic letter defines conditions that would be acceptable to utilize temporary non-ASME code repairs with NRC approval. The ASME recognized that relatively small flaws could remain in service without risk to the structural integrity of a piping system and developed ASME Code Case N-513. The Generic Letter 90-05 moderate energy limitations of 200°F and 275 psig for moderate energy piping were retained in the code case to maintain consistency with service conditions previously acceptable to the NRC as defined in Generic Letter 90-05. NRC approval of N-513 versions in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division l ," allows acceptance of partial through-wall or through-wall leaks for an operating cycle provided all conditions of the code case and NRC conditions are met. The code case also requires the Owner to demonstrate system operability due to leakage.

ASME Code Case N-513-3 provides analytical methods to be used for evaluating degraded piping conditions for determining structural integrity. The analytical methods provided in the code case are based on ASME Section XI, Appendix C, "Evaluation of Flaws in Piping," with supplemental guidance given in the code case specific to through-wall flaws. Linear Elastic Fracture Mechanics (LEFM) principles for evaluation of flaws in ferritic piping are normally employed. The ASME Section XI piping flaw evaluation methods do not place pressure or temperature limits for evaluating flaws in piping. The code case also allows evaluation by the branch reinforcement approach to allow evaluation of nonplanar through-wall flaws. The code case analytical methods account for flaw length, depth, pipe material toughness, applied stresses, and use of safety factors.

These analytical methods do not have a technical basis for limiting use to 275 psig, and would, in fact, be technically appropriate without a pressure limitation.

EGC has worked with a vendor to better understand the background, history, and effects of using ASME Code Case N-513-3 at a pressure of 375 psig in lieu of the current 275 psig limitation provided in the code case. The review identified that the NRC has

10CFR50.55a Relief Request ISR-08 (Page 3 of 7) previously granted relief for leaks on specific systems operating at temperatures greater than 200°F (see Enclosure 3 to the Reference 4 letter). EGC is seeking relief for general application for limited degradation in HPSW System raw water piping for a pressure 100 psig greater than the currently approved 275 psig. Raw water piping degradation is a well understood phenomenon and the evaluation methods in ASME Code Case N-513-3 are widely applied by the industry in raw water piping systems that operate at a pressure less than or equal to 275 psig without incident.

The structural aspects of raising the allowable operating pressure to 375 psig were evaluated as discussed in Enclosure 3 to the Reference 4 letter. It was determined that the code case allowable flaw sizes by both the LEFM and branch reinforcement methods used in ASME Code Case N-513-3 were smaller as would be expected. The effects of jet thrust force were evaluated and it was determined there was little difference in force for a 0.56" diameter flaw size at 275 psig versus 375 psig. The study also determined thatjet thrust force increases with increasing leakage rate and that it is appropriate to limit the application of this relief request to 375 psig. to the Reference 4 letter provides:

1. A review of relevant NRC approved relief requests

2. A structural integrity evaluation that includes:

- Design minimum wall thickness comparison

- ASME Code Case N-513-3 allowable flaw size comparison

- ASME Code Case N-513-3 cover thickness requirement comparison

3. A jet thrust force evaluation ASME Code Case N-513-3 requires that the Owner demonstrate system operability due to leakage. The code case does not demonstrate the consequence of leakage so the Owner is required to demonstrate leakage consequence/operability per operability procedures. This evaluation is demonstrated in an Operability Evaluation via EGC Procedure OP-AA-108-115, "Operability Determinations." The current licensing basis requirements and commitments, including the Technical Specifications and Updated Final Safety Analysis Report, are reviewed to establish the conditions and performance requirements to be met for determining operability, as necessary. The scope of an Operability Evaluation needs to be sufficient to address the capability of the System, Structure, and Component (SSC) to perform its specified safety function(s) from both the ASME Code Case N-513-3 structural perspective and leakage perspective. An Operability Evaluation should address the following, as applicable:

Determine what SSC is degraded, nonconforming, or unanalyzed.

Determine the extent of condition for all similarly affected SSC.

Determine the specified safety function(s) performed by the SSC.

Determine the circumstances of the potential nonconformance, including the possible failure mechanism.

Determine if the potential failure is time dependent and whether the condition will continue to degrade and/or will the potential consequences increase.

Determine the requirement or commitment established for the SSC, and why the requirement or commitment may not be met.

10CFR50.55a Relief Request ISR-08 (Page 4of7)

Determine by what means and when the potentially nonconforming SSC was first discovered.

Determine the basis for declaring the affected SSC operable, through:

o analysis, o test or partial test, o operating experience, and/or, o engineering judgment The HPSW System is a safety-related, open loop cooling water system at Peach Bottom Atomic Power Station. The primary function of the HPSW System is to provide cooling water flow, at a pressure greater than the Residual Heat Removal (RHR) System pressure, for removing heat from the RHR heat exchangers. The functional capability for the HPSW pumps is to provide a pressure at the heat exchanger service water outlet greater than the maximum RHR inlet pressure in the containment cooling mode. The HPSW System transfers heat from the RHR System to the service water system during operation in the following plant conditions:

Normal shutdown Post-accident shutdown Hot standby Refueling Normal plant operation The HPSW System is designed to:

Support post-accident containment heat removal Meet seismic Class I criteria Have sufficient capacity and redundancy to perform its safety-related functions Be operable during loss of offsite power Designed to ANSI B31.1, 1967 Edition The HPSW System at Peach Bottom Atomic Power Station, Units 2 and 3 has exhibited a history of degradation similar to raw fresh water systems throughout the nuclear industry.

Degradation requiring immediate action to address leakage or observed thinning in the system is generally due to localized corrosion mechanisms.

High Pressure Service Water System Description The major flow paths of the HPSW System consist of two independent parallel flow loops serving each unit. Each flow loop contains two HPSW pumps which discharge to a common header serving two RHR heat exchangers, connected in parallel, and then discharging through a pipe which is common to both loops. The HPSW pumps take suction from the Conowingo Pond through the Service Water Pump Bay and the HPSW loops discharge through a common pipe for each unit to the discharge pond. The discharge pipe contains a normally open motor-operated isolation valve and a pipe connection to the Emergency Cooling Water (ECW) System to provide an alternate discharge in the unlikely event that the Conowingo Dam fails or the pond floods. When

10CFR50.55a Relief Request ISR-08 (Page 5 of 7) the alternate discharge is used, the ECW System serves as a supply to the HPSW pumps through the pump bay. See Figure I5R-08-1 for a flow diagram of the Peach Bottom Atomic Power Station HPSW System.

A cross connection line connecting the two HPSW loops on each unit is provided including a normally closed motor-operated isolation valve. A cross connection line with two normally closed manual isolation valves is also provided between one Unit 2 HPSW loop and one Unit 3 HPSW loop. The cross connection lines provide the flexibility to establish alternate flow alignments if required under emergency conditions. A supply connection from the HPSW System to the RHR System, through two normally closed motor-operated valves, is provided from one HPSW loop per unit to permit the HPSW System furnishing a backup water supply to RHR for containment flooding. This is also known as Ultimate Cooling. The RHR and HPSW Systems are designed such that HPSW operates at a higher pressure than RHR; however, during standby conditions the RHR System pressure is maintained greater than HPSW. The RHR and HPSW Systems are standby systems that typically operate during testing or plant shutdown. Under this design, if there is an internal leak within a RHR heat exchanger, RHR water, which is normally torus water, leaks into the HPSW System and is discharged into the Conowingo Pond until the HPSW System pressure exceeds that of the RHR System.

Cross System Leakage Monitoring Each RHR heat exchanger contains a tube-to-shell differential pressure alarm, which is the first indication that there is an internal leak resulting in cross contamination from the RHR System to the HPSW System. Additionally, there are radiation monitors installed downstream of the HPSW System that indicate if there is cross system leakage. Between these alarms and established Operations and Chemistry procedures, the system is maintained such that unacceptable RHR System leakage into the HPSW System does not occur. HPSW piping through-wall leaks in an operating HPSW train would not contain unacceptable levels of radionuclides due to the actions described above to address system cross contamination and maintaining the HPSW System at a higher operating pressure than the RHR System. These actions assure any HPSW piping through-wall leaks would not result in an increase in the probability of release of radionuclides to the environment.

Summary EGC will apply ASME Code Case N-513-3 and Regulatory Guide 1.147, Revision 18 (Reference 3) (or later NRC defined revision as applicable) for evaluation of HPSW piping flaws at Peach Bottom Atomic Power Station if ASME code repairs cannot reasonably be completed within the Technical Specification required time limit. Peach Bottom Atomic Power Station will apply a 375 psig maximum operating pressure in lieu of the 275 psig maximum operating pressure defined in paragraph l(b) of the code case.

In addition, Peach Bottom Atomic Power Station will apply a 5 gpm leak limit to this relief request to limit the effects of jet thrust force even when evaluation of leakage effects would allow a higher leakage rate. Any leakage, if present, will be limited to the leakage allowed by the evaluation or 5 gpm, whichever is lower. The alternative retains

10CFR50.55a Relief Request ISR-08 (Page 6 of 7) acceptable structural and leakage integrity as described in the above paragraphs and would avoid the additional personnel radiation exposure and increase in plant risk associated with an unnecessary plant shutdown.

6. Duration of Proposed Alternative Relief is requested for the fifth ISI interval for Peach Bottom Atomic Power Station, Units 2 and 3. A Section XI compliant repair/replacement will be completed prior to exceeding the allowable period defined in Code Case N-513-3 Section l(e) and Regulatory Guide 1.147 or the next refueling outage, whichever comes first.

7. Precedents Peach Bottom Atomic Power Station, Units 2 and 3, fourth ISI interval Relief Request I4R-55 was authorized by NRC Safety Evaluation (SE) dated March 19, 2015 (Reference 5). Relief Request I5R-08 for the Peach Bottom Atomic Power Station, Units 2 and 3, fifth ten-year ISI interval, utilizes a similar approach to the previously approved relief request.

8. References

1. ASME Code Case N-513-3, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division l," Cases of the ASME Boiler and Pressure Vessel Code, dated January 26, 2009.

2. NRC Generic Letter 90-05, "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping," dated June 15, 1990.

3. NRC Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 18.

4. Letter from J. Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information -

Proposed Alternative to Utilize Code Case N-513-3, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division l," at a Higher System Operating Pressure," dated February 5, 2015.

5. Letter from T. L. Tate (U.S. Nuclear Regulatory Commission) to B. C. Hanson (Exelon Generation Company, LLC), "Peach Bottom Atomic Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2 - Relief from the Requirements of the ASME Code (TAC Nos. MF3799, MF3800, MF3801, and MF3802)," dated March 19, 2015 (ADAMS Accession No. ML15043A496).

10CFRSO.SSa Relief ISR-08 Revision 0 (Page 7of7)

Figure ISR-08-1 Peach Bottom Atomic Power Station HPSW Flow Diagram

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