NL-17-1713, Proposed Alternative GEN-ISI-ALT-2017-03, and HNP-ISI-ALT-05-07, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Moderate Pressure, and for Higher Pressure

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Proposed Alternative GEN-ISI-ALT-2017-03, and HNP-ISI-ALT-05-07, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Moderate Pressure, and for Higher Pressure
ML18096B554
Person / Time
Site: Hatch, Farley  Southern Nuclear icon.png
Issue date: 04/06/2018
From: Gayheart C
Southern Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-17-1713
Download: ML18096B554 (13)
v • d • e


Text

.t. Southern Nuclear Regulatory Affairs 40 inverness Center Parkway Post Office Box 1295 Birmingham, AL 35242 205 992 5000 tel APR 0 6 2018 205 992 7601 fax Docket Nos.: 50-321 50-348 NL-17-1713 50-366 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant- Unit 1 and 2 Joseph M. Farley Nuclear Plant- Unit 1 and 2 Proposed Alternative GEN-ISI-ALT-2017-03, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Moderate Pressure Edwin I. Hatch Nuclear Plant- Unit 1 and 2 Proposed Alternative HNP-ISI-ALT-05-07, Version 1.0 Service Water Evaluation for Code Case N-513-4 for Higher Pressure Ladies and Gentlemen:

In accordance with the provisions of 10 CFR 50.55a(z)(2}, Southern Nuclear Operating Company (SNC) hereby requests Nuclear Regulatory Commission (NRC) approval of proposed inservice inspection (lSI) alternative GEN-ISI-ALT-2017-03, Version 1. For Farley Nuclear Plant Units 1 and 2 and Hatch Nuclear Plant Units 1 and 2, this request proposes an alternative from the requirement to perform repair/replacement activities for moderately 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.

SNC is also requests NRC approval of proposed lSI alternative HNP-ISI-ALT-05-07, Version 1.

For Hatch Nuclear Plant Units 1 and 2, this request proposes an alternative from the requirement to perform repair/replacement activities for moderately degraded Class 2 and 3 piping whose maximum operating temperature does not exceed 200°F and whose maximum operating pressure does not exceed 450 psig. Specifically, this applies to the Residual Heat Removal Service Water (RHRSW) system.

SNC is requesting approval to apply the evaluation methoqs of ASME Code Case N-513-4, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1,"to Class 2 and 3 components that meet the operational and configuration limitations of Code Case N-513-4, paragraphs 1 (a}, 1 (b), 1 (c), and 1 (d) 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.

These proposed alternatives are being submitted in accordance with 10 CFR 50.55a(z)(2),

"hardship without a compensating increase in the level of quality and safety" to allow SNC to

U. S. Nuclear Regulatory Commission NL-17-1713 Page 2 perform additional extent of condition examinations on the affected systems while allowing time for safe and orderly long-term repair actions if necessary.

SNC requests NRC review and approval of this alternative by April1, 2019.

This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at 205.992.7369.

Respectfully submitted, eart Regulatory Affairs Director CAG/NDJ

Enclosures:

1. Proposed Alternative GEN-ISI-ALT-2017-03, Version 1.0 in accordance with 10 CFR 50.55a(z)(2) for Moderate Pressure
2. Proposed Alternative HNP-ISI-ALT 07, Version 1.0 in accordance with 10 CFR 50.55a(z)(2) for Higher Pressure Cc: Regional Administrator, Region II NRR Project Manager- Farley Nuclear Plant NRR Project Manager- Hatch Nuclear Plant Senior Resident Inspector- Farley Nuclear Plant Senior Resident Inspector- Hatch Nuclear Plant RTYPE: CFA04.054, CHA02.004

Edwin I. Hatch Nuclear Plant Unit 1 and 2 Joseph M. Farley Nuclear Plant Unit 1 and 2 Enclosure 1 Proposed Alternative GEN-ISI-ALT-2017-03, Version 1.0 in accordance with 10 CFR 50.55a(z){2) for Moderate Pressure

Enclosure 1 to NL-17-1713 Proposed Alternative GEN-ISI-ALT-2017-03, Version 1.0 in accordance with 10 CFR 50.55a(z)(2) for Moderate Pressure Hatch Nuclear Plant and Farley Nuclear Plant Proposed Alternative in Accordance with 10 CFR 50.55a(z){2)

Hardship without a Compensating Increase in Quality and Safety

1. ASME Code Component(s) Affected All American Society of Mechanical Engineers (ASME),Section XI, Class 2 and 3 components that meet the operational and configuration limitations of Code Case N-513-4, paragraphs 1 (a),

1 (b), 1 (c), and 1 (d).

2. Applicable Code Edition and Addenda

The following table identifies the ASME Section XI Code of Record for performing lnservice Inspection (lSI) activities at Southern Nuclear Plant Hatch and Plant Farley.

ASME Section XI Interval lSI Interval Plant Interval Edition/Addenda End Start (scheduled)

Hatch Nuclear Plant, 2007 Edition/2008 5 01/01/2016 12/31/2025 Units 1 and 2 Addenda Farley Nuclear Plant, 2007 Edition/2008 5 12/1/2017 11/30/2027 Units 1 and 2 Addenda

3. Applicable Code Requirements ASME Code,Section XI, IWC-3120 and IWC-3130 require that flaws exceeding the defined acceptance criteria be corrected by repair/replacement activities or evaluated and accepted by analytical evaluation. ASME Code,Section XI, IWD-3120(b) requires that components exceeding the acceptance standards of IWD-3400 be subject to supplemental examination, or a repair/replacement activity.

4. Reason for Request

In accordance with 10 CFR 50.55a(z)(2), Southern Nuclear Company (SNC) 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. 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 SNC 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 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 does not allow evaluation of flaws located away from attaching circumferential piping welds that are in elbows, bent pipe, reducers, expanders, and branch E1-1

Enclosure 1 to NL-17-1713 Proposed Alternative GEN-ISI-ALT-2017-03, Version 1.0 in accordance with 10 CFR 50.55a(z)(2) for Moderate Pressure tees. ASME Code Case N-513-3 also does not allow evaluation of flaws located in heat exchanger external tubing or piping. ASME Code Case N-513-4 provides guidance for evaluation of flaws in these locations.

5. Proposed Alternative and Basis for Use SNC is requesting approval to apply the evaluation methods of ASME Code Case N-513-4, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1," to Class 2 and 3 components that meet the operational and configuration limitations of Code Case N-513-4, paragraphs 1 (a), 1 (b), 1 (c), and 1 (d) 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 NRC issued Generic Letter 90-05 (Reference 1), "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping (Generic Letter 90-05)," to address the acceptability of limited degradation in piping with maximum operating conditions less than or equal to 200°F (93°C} and less than or equal to 275 psig (1.9 MPa). The generic letter defines conditions that would be acceptable to utilize temporary non-code repairs with NRC approval.

ASME recognized that relatively small flaws could remain in service without risk to the structural integrity of a piping system and developed Code Case N-513. The NRC approval of Code Case N-513-3 in Regulatory Guide 1.147 (Reference 2}, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1," 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 recognized that the limitations in Code Case 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. Code Case 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 3 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 3 letter provides the ASME approved Code Case N-513-4. The following provides a high level overview of the Code Case N-513-4 changes:

1) 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 (R 0 t) 112 from the centerline of the attaching circumferential piping weld.

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 Code Case N-513-4 when compared to NRC approved Code Case N-513-3 is provided in Attachment 4 of the Reference 3 letter.

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Enclosure 1 to NL-17-1713 Proposed Alternative GEN-ISI-ALT-2017-03, Version 1.0 in accordance with 10 CFR 50.55a(z}{2) for Moderate Pressure The design basis is considered for each leak and evaluated using the SNC Operability Evaluation process. The evaluation process must consider requirements or commitments established for the system, continued degradation and potential consequences, operating experience, and engineering judgment. 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 (ML14240A603, ML14316A167, ML15070A428). The periodic inspection interval defined using paragraph 2(e) of Code Case 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 1 (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 that can be tolerated relative to room flooding, among others. The safety factor of four (4) on leakage is based upon Code Case N-705 (Reference 4),

which is accepted without condition in Regulatory Guide 1.147, Revision 17. 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 nonplanar 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 Code Case N-705. Note that the alternative herein does not propose to use any portion of Code Case N-705 and that citation of N-705 is intended only to provide technical basis for the safety factor on leakage.

During the temporary acceptance period, leaking flaws will be monitored daily as required by paragraph 2(f) of Code Case 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, SNC will apply ASME Code Case N-513-4 to evaluation of Class 2 and 3 components that are within the scope of the Code Case. 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 concert 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 Code Case N-513-4 for Class 2 and Class 3 components within the scope of the Code Case. A Section XI compliant repair/replacement will be completed prior to exceeding the next refueling outage or allowable flaw size, whichever comes first. This relief request will be applied for the duration of the inservice inspection intervals defined in E1-3

Enclosure 1 to NL-17-1713 Proposed Alternative GEN-ISI-ALT-2017-03, Version 1.0 in accordance with 10 CFR 50.55a(z)(2) for Moderate Pressure Section 2 of this relief request or such time as the NRC approves Code Case N-513-4 in Regulatory Guide 1.147 or other document. If a flaw is evaluated near the end of the interval 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. Precedents The NRC has approved a similar generic relief request submitted by Exelon. There have also been several submittals approved for N-513-4 use in specific applications. The table below lists several Safety Evaluation Reports as precedents for use of Code Case N-513-4.

SER Accession No. Plant Application Additional Requirement ML16230A237 Exelon Fleet Generic N-513-4 Critical leakage determination ML15070A428 ANO Leaking sweepolet 5 gpm leakage limit ML14316A167 Fort Calhoun Leaking elbow None ML14335A551 Peach Bottom Leaking elbow 5 gpm leakage limit

8. References
1. 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)"
2. Regulatory Guide 1.147, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 17, August 2014.
3. Letter from D. Gudger (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "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 1 ",dated January 28, 2016 (ML16029A003).
4. 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 1,"October 12, 2006.

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Edwin I. Hatch Nuclear Plant Unit 1 and 2 Enclosure 2 Proposed Alternative HNP-151-ALT-05-07, Version 1.0 in accordance with 10 CFR 50.55a(z)(2) for Higher Pressure

Enclosure 2 to NL-17-1713 Proposed Alternative HNP-ISI-ALT-05-07, Version 1.0 in accordance with 10 CFR 50.55a(z}{2) for Higher Pressure

1. ASME Code Component(s) Affected All American Society of Mechanical Engineers (ASME),Section XI, Class 2 and 3 components that meet the operational and configuration limitations of Code Case N-513-4, paragraphs 1 (a),

1 (b), and 1 (d). This relief request applies to Hatch Nuclear Plant Unit 1 and Unit 2 Residual Heat Removal Service Water (RHRSW) system at a higher pressure limit of 450 psig instead of 275 psig noted in 1 (c). Maximum operating temperature does not exceed 200°F as noted in 1

~). .

2. Applicable Code Edition and Addenda

The following table identifies the ASME Section XI Code of Record for performing lnservice Inspection (lSI) activities at Southern Nuclear Plant Hatch.

lSI ~SME Section XI Interval Interval Plant Interval Edition/Addenda Start End Hatch Nuclear Plant, ~007 Edition/2008 5 01/01/2016 12/31/2025 Units 1 and 2 ~ddenda

3. Applicable Code Requirements ASME Code,Section XI, IWC-3120 and IWC-3130 require that flaws exceeding the defined acceptance criteria be corrected by repair/replacement activities or evaluated and accepted by analytical evaluation. ASME Code,Section XI, IWD-3120(b) requires that components exceeding the acceptance standards of IWD-3400 be subject to supplemental examination, or to a repair/replacement activity.

4. Reason for Request

In accordance with 10 CFR 50.55a(z)(2), Southern Nuclear Company (SNC) 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 450 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 SNC 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 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 does not allow evaluation of flaws located away from attaching circumferential piping welds that are in elbows, bent pipe, reducers, expanders, and branch tees. ASME Code Case N-513-3 also does not allow evaluation of flaws located in heat exchanger external tubing or piping. ASME Code Case N-513-4 provides guidance for evaluation of flaws in these locations.

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Enclosure 2 to NL-17-1713 Proposed Alternative HNP-ISI-ALT-05-07, Version 1.0 in accordance with 10 CFR 50.55a(z)(2) for Higher Pressure

5. Proposed Alternative and Basis for Use SNC is requesting approval to apply the evaluation methods of ASME Code Case N-513-4, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1,"to Class 2 and 3 components that meet the operational and configuration limitations of Code Case N-513-4, paragraphs 1 (a), 1 (b), and 1 (d) 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. RHRSW system at Hatch Unit 1 and Unit 2 has a maximum operating pressure of 450 psig. It is desired to expand the scope of N-513-4 such that it may be generically applied to this specific system. Maximum operating temperature does not exceed 200°F as noted in 1 (c).

The NRC issued Generic Letter 90-05 (Reference 1), "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping (Generic Letter 90-05)," to address the acceptability of limited degradation in piping. The generic letter defines conditions that would be acceptable to utilize temporary non-code repairs with NRC approval. ASME recognized that relatively small flaws could remain in service without risk to the structural integrity of a piping system and developed Code Case N-513. The NRC approval of Code Case N-513-3 in Regulatory Guide 1.147 (Reference 2), "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1,"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 recognized that the limitations in Code Case 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. Code Case 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 3 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 3 letter provides the ASME approved Code Case N-513-4. The following provides a high level overview of the Code Case N-513-4 changes:

1) 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 (R 0 t) 112 from the centerline of the attaching circumferential piping weld.
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 Code Case N-513-4 when compared to NRC approved Code Case N-513-3 is provided in Attachment 4 of the Reference 3 letter.

The design basis is considered for each leak and evaluated using the SNC Operability Evaluation process. The evaluation process must consider requirements or commitments established for the system, continued degradation and potential consequences, operating experience, and engineering judgment. As required by the Code Case, the evaluation process E2-2 to NL-17-1713 Proposed Alternative HNP-ISI-ALT-05-07, Version 1.0 in accordance with 10 CFR 50.55a(z)(2) for Higher Pressure 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 (ML14240A603, ML14316A167, ML15070A428). The periodic inspection interval defined using paragraph 2(e) of Code Case 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 1 (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 that can be tolerated relative to room flooding, among others. The safety factor of four (4) on leakage is based upon Code Case N-705 (Reference 4),

which is accepted without condition in Regulatory Guide 1.147, Revision 17. 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 nonplanar 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 Code Case N-705. Note that the alternative herein does not propose to use any portion of Code Case N-705 and that citation of N-705 is intended only to provide technical basis for the safety factor on leakage.

During the temporary acceptance period, leaking flaws will be monitored daily as required by paragraph 2(f) of Code Case 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.

The Class 3 RHRSW system at Hatch Unit 1 and Unit 2 has a maximum operating pressure of 450 psig per plant procedures (Reference 6). It is desired to expand the scope of N-513-4 such that it may be generically applied to this specific system. The discussion provided herein demonstrates that the use of Code Case N-513-4 with the higher pressure limit will reduce plant burden without any adverse effect on safety.

  • First, relevant NRC relief requests were reviewed as shown in Section 7. They demonstrate that there is precedent for the temporary acceptance of leaking flaws in Class 2/3 piping and components at pressures higher than 275 psig.
  • Second, a structural integrity evaluation (Reference 7) was performed to determine the impact of the increased pressure on design minimum wall thickness, N-513-4 allowable flaw sizes and the N-513-4 cover thickness requirement. While the influence of the higher pressure was observed in the evaluation, structural integrity can still be demonstrated. The functionality and validity of the Code Case methods at higher pressure were confirmed.
  • Finally, jet thrust forces were estimated (Reference 7) for a leaking pipe at 275 and 450 psig.

A significant change in jet thrust forces was only seen with large opening areas that would E2-3

Enclosure 2 to NL-17-1713 Proposed Alternative HNP-ISI-ALT-05-07, Version 1.0 in accordance with 10 CFR 50.55a(z)(2) for Higher Pressure result in high leak rates, i.e., rates that would challenge system functionality or local spray and/or compartment flooding requirements.

Furthermore, the following additional requirements are included as part of the Relief Request for usage on higher pressure RHR system to bolster defense-in-depth and avoid adverse consequences:

requirement for high energy Class 3 applications.

  • The leakage shall be stopped throughout the temporary acceptance period using engineered mechanical clamping designed by Hatch. The engineered mechanical clamping design shall be sufficient to withstand the maximum operating pressure and removable such that the frequent periodic inspections defined in paragraph 2(e) of N-513-4 may be performed.

In summary, SNC will apply ASME Code Case N-513-4 to evaluation of Class 2 and 3 components that are within the scope of the Code Case except for a higher pressure limit of 450 psig instead of 275 psig noted in 1 (c). 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 concert with safety factors on leakage limits, expanding the number of augmented inspection locations from 5 to 10, and requiring the leakage be stopped during the temporary acceptance period 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 Code Case N-513-4 for Class 2 and Class 3 components within the scope of the Code Case with pressure limit of 450 psig. A Section XI compliant repair/replacement will be completed prior to exceeding the next refueling outage or allowable flaw size, whichever comes first. This relief request will be applied for the duration of the inservice inspection intervals defined in Section 2 of this relief request or such time as the NRC approves Code Case N-513-4 in Regulatory Guide 1.14 7 or other document. If a flaw is evaluated near the end of the interval 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. Precedents The NRC has approved a similar generic relief request submitted by Exelon. There have also been several submittals approved for N-513-4 use in specific applications. The table below lists several Safety Evaluation Reports as precedents for use of Code Case N-513-4.

E2-4

Enclosure 2 to NL-17-1713 Proposed Alternative HNP-ISI-ALT-05-07, Version 1.0 in accordance with 10 CFR 50.55a(z)(2) for Higher Pressure SER Accession No. Plant Application Additional Requirement ML16230A237 Exelon Fleet Generic N-513-4 Critical leakage determination ML15070A428 ANO Leaking sweepolet 5 gpm leakage limit ML14316A167 Fort Calhoun Leaking elbow None ML14335A551 Peach Bottom Leaking elbow 5 gpm leakage limit A search of the NRC ADAMS database was conducted to identify any relief requests for continued operation of degraded high energy piping or components. The search results are summarized below.

ADAMS Accession Plant I Operating Number Condition Description Status McGuire, U1 I Application for continued operation of ML072140851 2,500 psig leaking valve Approved San Onofre U2, Generic application for continued U3 I < 275 psig operation of high temperature through-ML101440381 (275° F) wall leaking pipe Approved Exelon Plants I Generic request to use Code Case N-ML15043A496 375 psig 513-3 at a higher pressure limit Approved

8. References
1. 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)"
2. Regulatory Guide 1.147, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1,"Revision 17, August 2014.
3. Letter from D. Gudger (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "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 1 ","dated January 28, 2016(ML16029A003).
4. 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 1,"October 12, 2006.
5. NRC Standard Review Plan, NUREG-0800, Branch Technical Position 3-3, "Protection Against Postulated Piping Failures in Fluid Systems Outside Containment," Revision 3, March 2007.
6. Southern Nuclear Operating Procedures, "Residual Heat Removal System," 34SO-E11-01 0-1, Version 44.12. (Hatch Unit 1) and 34SO-E11-01 0-2, Version 42.9 (Hatch Unit 2).
7. Structural Integrity Associates, Inc, Technical Basis for Proposed Alternative to Use ASME Code Case N-513-4 and Scope Expansion to a Higher Pressure Limit," January 20, 2017, Report No. 1601270.402.RO.

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