NL-21-0989, License Amendment Request to Relocate Augmented Piping Inspection Program Details

License Amendment Request to Relocate Augmented Piping Inspection Program Details

ML21356B484
Person / Time
Site:	Vogtle, Farley
Issue date:	12/22/2021
From:	Gayheart C Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-21-0989
Download: ML21356B484 (21)

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Southern Nuclear Operating Company Joseph M. Farley Nuclear Power Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Relocate Augmented Piping Inspection Program Details from Technical Specifications to a Licensee Controlled Document Enclosure Evaluation of Proposed Change

Evaluation of Proposed Change E-1 1.0

SUMMARY

DESCRIPTION Southern Nuclear Operating Company (SNC) hereby requests a license amendment to Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2 renewed facility operating licenses NPF-2 and NPF-8, respectively, and Vogtle Electric Generating Plant (VEGP) Units 1 and 2 renewed facility operating licenses NPF-68 and NPF-81, respectively.

The proposed change will revise the FNP Technical Specifications (TS) by relocating some detailed information from TS 5.5.16, Main Steamline Inspection Program, to the FNP Updated Final Safety Analysis Report (UFSAR). The proposed change also revises the VEGP TS by relocating some detailed information from TS 5.5.16, MS and FW Piping Inspection Program, to the VEGP UFSAR. A program description will remain in TS 5.5.16 for each plant. The UFSARs for each plant will be updated in accordance with 10 CFR 50.71(e).

The proposed changes are requested to remove detail from the TS and to provide a consistent program description for this augmented inspection program. To determine if information in these TS can be relocated to licensee control, these changes are compared to the criteria in 10 CFR 50.36(c) with regard to what items are contained in TS.

2.0 DETAILED DESCRIPTION 2.1 Regulatory Basis TS 5.5.16 for each plant provides details of an augmented inservice inspection program for high energy lines outside of containment. These requirements apply to welds in piping systems or portions of piping systems located outside of containment where protection from the consequences of postulated ruptures is not provided by a system of pipe whip restraints, jet impingement barriers, protective enclosures and/or other measures designed specifically to cope with such ruptures. These augmented inservice inspections provide assurance of the continued integrity of the piping systems over their service life. These requirements are contained in the UFSARs for both FNP and VEGP. The requirements were addressed during initial licensing of both plants as described below.

Farley The main steam line break for FNP is not postulated for main steam piping outside containment up to and including the main steam isolation valves. The basis for this position is discussed in UFSAR Appendix 3K, which describes the criteria for postulating pipe ruptures or cracks in high-energy lines outside containment and the methods for evaluating the effects of these breaks.

Attachment A, Part II, of Appendix 3K specifically addresses the postulated break and leakage locations in the main steam lines outside of containment and demonstrates that this piping conforms to Branch Technical Positions (BTP) APCSB 3-1 and MEB 3-1 of Standard Review Plan (SRP) Sections 3.6.1 and 3.6.2, respectively. This is consistent with Section 3.6 of Supplement 1 of the NRC Safety Evaluation Report for the Joseph M. Farley Nuclear Plant -

Units 1 and 2 (Reference 1).

As described in UFSAR Appendix 3K, Part II, Section D, inservice examination and related design provisions in the containment penetration area and throughout the no break region should address location of protective measures and scope of the augmented inspection. The protective measures, structures, and guard pipes should not prevent the access required to conduct the inservice examinations specified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Division 1, Rules for Inspection and

Evaluation of Proposed Change E-2 2.2 Current Technical Specification Requirements Current TS 5.5.16 FNP Units 1 and 2 5.5.16 Main Steamline Inspection Program The three main steamlines from the rigid anchor points of the containment penetrations downstream to and including the main steam header shall be inspected. The extent of the inservice examinations completed during each inspection interval (IWA 2400, ASME Code, 1974 Edition,Section XI) shall provide 100 percent volumetric examination of circumferential and longitudinal pipe welds to the extent practical. The areas subject to examination are those defined in accordance with examination category C-G for Class 2 piping welds in Table IWC-2520.

Testing of Components in Light-Water Cooled Plants. For those portions of fluid system piping identified in B.2.c of APCSB 3-1, the extent of inservice examinations completed during each inspection interval should provide 100-percent volumetric examination of circumferential and longitudinal pipe welds within the boundary of these portions of piping to the extent practical.

Vogtle As described in the Vogtle UFSAR, Section 3.6.1, the selection of pipe failure locations and evaluation of the consequences on nearby essential systems, components, and structures are in accordance with the requirements of 10 CFR 50, Appendix A, General Design Criterion, and BTPs ASB 3-1 and MEB 3-1.

The Nuclear Regulatory Commission (NRC) reviewed and addressed the selection of pipe failure locations in the NRC Safety Evaluation Report for Vogtle (Reference 2). Section 3 describes that the main steam and feedwater systems up to the first restraint outside containment are classified as part of the break exclusion boundary as defined in item B.1.6 of BTP MEB 3-1, Postulated Breaks and Leakage Locations in Fluid System Piping Outside Containment. For high energy piping within the containment penetration area where breaks are not postulated, SRP Section 3.6.2 sets forth certain criteria for the analysis and subsequent augmented inservice inspection requirements. Breaks need not be postulated in those portions of piping within the containment penetration region that meet the requirements of ASME Code Section III, Subarticle NE-1120 and the additional requirements outlined in BTP MEB 3-1 of SRP Section 3.6.2. Augmented inservice inspection is required for those portions of piping within the break exclusion region.

UFSAR Section 6.6.8 describes the augmented inservice inspection program for high-energy fluid systems piping between containment isolation valves or where no isolation valve is used inside containment, between the first rigid pipe connection to the containment penetration or the first pipe whip restraint inside containment and the outside isolation valve. This program includes 100 percent volumetric examination of circumferential and longitudinal pipe welds in the affected piping during each inspection interval and is conducted in accordance with ASME Code,Section XI, and covers the high-energy fluid systems described in UFSAR subsections 3.6.1 and 3.6.2.

Evaluation of Proposed Change E-3 VEGP Units 1 and 2 5.5.16 MS and FW Piping Inspection Program This program shall provide for the inspection of the four Main Steam and Feedwater lines from the containment penetration flued head outboard welds, up to the first five-way restraint. The extent of the inservice examinations completed during each inspection interval (ASME Code Section XI) shall provide 100% volumetric examination of circumferential and longitudinal welds to the extent practical. This augmented inservice inspection is consistent with the requirements of NRC Branch Technical Position MEB 3-1, "Postulated Break and Leakage Locations in Fluid System Piping Outside Containment,"

November 1975 and Section 6.6 of the FSAR.

2.3 Reason for the Proposed Change The proposed changes are requested to relocate detail from the TS to the plants UFSARs and to provide a consistent program description for this augmented inspection program. The TS details proposed for relocation from the TS are descriptions of inspections to be performed on plant equipment. These details are design basis information and are more appropriately contained in the respective plants UFSARs. Therefore, SNC requests relocation of these TS details to the respective plants UFSARs as described below.

2.4 Description of the Proposed Change Proposed TS Change FNP Units 1 and 2 5.5.16 Main Steamline Inspection Program In accordance with the augmented inservice inspection program for high energy lines outside of containment, examinations of welds in the main steam lines of each unit shall be performed to provide assurance of the continued integrity of the piping systems over their service lifetime. These requirements apply to welds in piping systems or portions of systems located outside of containment where protection from the consequences of postulated ruptures is not provided by a system of pipe whip restraints, jet impingement barriers, protective enclosures and/or other measures designed specifically to cope with such ruptures.

VEGP Units 1 and 2 5.5.16 MS and FW Piping Inspection Program In accordance with the augmented inservice inspection program for high energy lines outside of containment, examinations of welds in the main steam and main feedwater lines of each unit shall be performed to provide assurance of the continued integrity of the piping systems over their service lifetime. These requirements apply to welds in piping systems or portions of systems located outside of containment where protection from the consequences of postulated ruptures is not provided by a system of pipe whip restraints, jet impingement barriers, protective enclosures and/or other measures designed specifically to cope with such ruptures.

The current detailed program description contained in TS 5.5.16 will be relocated to the UFSARs for FNP and VEGP. The requirements in the current TS 5.5.16 for each plant are

Evaluation of Proposed Change E-4 already contained in each plants respective UFSAR. However, the specific verbiage of TS 5.5.16 will be included in each plants UFSAR as a concise description of the augmented inspection program. Specifically, the entire verbiage of the FNP TS 5.5.16 will be relocated to the FNP UFSAR, Section 3K, Part II, Item D. Likewise, the entire verbiage of the VEGP TS 5.5.16 will be relocated to the VEGP UFSAR, Section 6.6.8. Any UFSAR information that duplicates the relocated TS information may be edited for clarity. The UFSARs for each plant will be updated in accordance with 10 CFR 50.71(e).

No changes are proposed for any Limiting Condition for Operation in Section 3.0 of the FNP or VEGP TS in this request. Therefore, this license amendment request is not linked to the previously submitted VEGP license amendment request to remove one Main Steam Isolation Valve System from TS 3.7.2 (ML21274A073).

3.0 TECHNICAL EVALUATION

Detailed discussion associated with FNP TS 5.5.16, Main Steamline Inspection Program, and VEGP TS 5.5.16, MS and FW Piping Inspection Program, are proposed to be relocated to the licensee controlled UFSARs as described above. These detailed descriptions are replaced with program descriptions in TS 5.5.16 for each plant. The program description is contained in each plants UFSAR. The licensing basis for the augmented inservice inspection program of high energy piping outside of containment is not changed by the relocation of details about the inspection program from the TS to the UFSAR.

The FNP and VEGP requirements for the augmented inspection program for high energy lines from the containment wall to and including the inboard or outboard isolation valves (per MEB 3-1, ASP 3-1 and APCSB 3-1) are consistent with the requirements of BTP 3-4. BTP 3-4 is used below for easier reference.

10 CFR 50.36 Criteria The augmented inspection program described in TS 5.5.16 is one of several design criteria provided by the NRC in BTP 3-4 to establish a pipe break exclusion zone for ASME Class 2 secondary system high energy piping. This inspection is in addition to the inservice inspections of piping required by 10 CFR 50.55a. The augmented inspection is only applied to the secondary system piping for which plant designers did not provide barriers, pipe whip restraints, or enclosures to protect against the effects of a postulated secondary system pipe rupture.

These sections of piping are described in each plants UFSAR.

The NRC has provided guidance for the contents of TS in 10 CFR 50.36(c). As a result, TS requirements which fall within or satisfy any of the criteria in the regulation must be retained in the TS, while those requirements which do not fall within or satisfy these criteria may be relocated to licensee-controlled documents.

The regulatory requirements are listed below.

(1) Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity.

Evaluation of Proposed Change E-5 The details in TS 5.5.16 which are being relocated to the facility UFSARs are not safety limits or limiting safety system settings. They describe an augmented inspection program for certain secondary system high energy piping. This proposed TS change does not meet the Safety Limits criterion.

(2) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

Criterion 1 - Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

The details in TS 5.5.16, which are being relocated to the facility UFSARs, are not applicable to installed instrumentation which is used to detect, and indicate in the control room, an abnormal degradation of the reactor coolant pressure boundary. It describes an augmented inspection program for certain secondary system high energy piping. TS 5.5.16 does not address instrumentation and does not impact the reactor coolant boundary. No physical change to the plant or the piping inspection program is proposed by this request. This TS does not meet Criterion 1.

Criterion 2 - A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The details of TS 5.5.16, which are being relocated to the facility UFSARs, are not applicable to a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The details describe an augmented inspection program for certain secondary system high energy piping. No physical change to the plant or the piping inspection program is proposed by this request. Therefore, this TS does not meet Criterion 2.

Criterion 3 - A structure, system or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Per BTP 3-4, the augmented inspection program described in TS 5.5.16 is part of the design criteria used to determine the postulated break locations in certain secondary system high energy lines in the power plant. The TS describes a piping inspection program. It does not describe a structure, system, or component which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. No physical change to the plant or the piping inspection program is proposed by this request. Therefore, this TS does not satisfy criterion 3.

Criterion 4 - A structure, system or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

No specific TS-related structure, system or component is being revised or removed from the TS.

Per BTP 3-4, the augmented inspection program described in TS 5.5.16 is part of the criteria used to determine if pipe breaks must be postulated in certain secondary system high energy lines in the power plant. This design criteria, including the augmented inspection program,

Evaluation of Proposed Change E-6 (7) Initial notifications are made to the NRC Operations Center (8) Written Reports. Licensees must submit Licensee Event Reports in accordance with 10 CFR 50.73 when required.

These criteria do not apply to the relocation of information from the TS to the UFSAR.

As discussed above, the details contained in TS 5.5.16 do not satisfy the 10 CFR 50.36(c) categories and criteria for inclusion of information in TS. Thus, relocation of the detailed information from TS 5.5.16 to the UFSAR of each plant is appropriate.

establishes that certain secondary system pipe breaks need not be assumed, and design provisions are not needed to mitigate them. Therefore, this TS does not meet Criterion 4.

(3) Surveillance requirements. Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

The inspections described in TS 5.5.16 do not meet the requirements for a SR because they are not related to the test, calibration or inspection to assure the necessary quality of systems and components. The inservice inspections needed to assure the necessary quality of systems and components are required by 10 CFR 50.55a. The augmented inspections are performed in addition to the ASME Code Section XI inspections/examinations required by 10 CFR 50.55a.

Both the required inservice inspection and the augmented inservice inspection will continue to be performed in accordance with the requirements in each plants UFSAR. Additionally, the NRC Policy Statement on Technical Specifications (Reference 3) indicates that SRs are associated with Limiting Conditions for Operation. There is no Limiting Condition for Operation related to this proposed change. Therefore, these details do not qualify as a SR in the TS.

(4) Design features. Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c) (1),

(2), and (3) of this section.

The information in TS 5.5.16 augmented inservice inspection program for high energy lines outside of containment does not describe a facility design feature, such as materials of construction and geometric arrangements, that could have a significant effect on safety.

(5) Administrative controls. Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in § 50.4.

TS 5.5.16 is not an administrative control relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure safe operation. However, it is retained in the Administrative Controls section. The program details are relocated to each plants UFSAR.

(6) Decommissioning applies only to nuclear power plants that have submitted decommissioning certifications

Evaluation of Proposed Change E-7

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent The Surry Power Station license amendment request listed below requested the relocation of two TS containing augmented inspection requirements from the TS to their Technical Requirements Manual (TRM). In addition, they added an augmented inspection program description to the Administrative Controls section of their TS. The program description added to the Administrative Controls provided a description of the augmented inspection program that did not include detailed information about how the program requirements were met.

Letter from Virginia Electric and Power Company to NRC, Proposed License Amendment Request, Relocation of Augmented Inspection Requirements from TS 4.2 and TS 4.15 to Technical Requirements Manual, dated April 11, 2014 (ML14112A073)

Letter from NRC to Virginia Electric Power Company, Surry Power Station Unit Nos 1 and 2, Issuance of Amendments Regarding Relocation of Augmented Inspection Requirements, dated April 28, 2015 (ML15099A679)

The following NRC requirements and guidance documents are applicable to the review of the proposed changes.

10 CFR 50.36(c) provides the requirements for items that must be contained in plant Technical Specifications. As described above the details associated with the augmented inspection program for high energy piping outside containment do not meet the criteria set forth in 10 CFR 50.36.

10 CFR 50.55a(g)(6)(ii) states that, "The Commission may require the licensee to follow an augmented inservice inspection program for systems and components for which the Commission deems that added assurance of structural reliability is necessary." The augmented inspections are performed in addition to required ASME Code Section XI inspections or examinations and will continue to be performed as required by the UFSARs for each plant.

10 CFR 50, Appendix A, GDC 4, "Environmental and dynamic effects design bases," provides requirements for certain systems to be able to withstand postulated events without loss of function. Compliance with GDC 4 is not impacted by the relocation of details from the TS to the UFSAR for augmented inservice inspection of high energy piping outside containment because the design basis for the affected piping is not changed by the relocation of inspection details to the UFSAR.

The proposed TS changes are consistent with and in compliance with the above regulatory requirements and criteria. Therefore, the proposed changes will assure safe operation by continuing to meet applicable regulations and requirements.

Evaluation of Proposed Change E-8 4.3 No Significant Hazard Consideration Determination Analysis Southern Nuclear Operating Company (SNC) has evaluated the proposed changes to the Technical Specifications (TS) using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The proposed change will revise the Farley Nuclear Plant (FNP) TS by relocating some detailed information from TS 5.5.16, Main Steamline Inspection Program, to the FNP Updated Final Safety Analysis Report (UFSAR). The proposed change also revises the Vogtle Electric Generating Plant (VEGP) TS by relocating some detailed information from TS 5.5.16, MS and FW Piping Inspection Program, to the VEGP UFSAR. A program description will remain in TS 5.5.16 for both plants.

As required by 10 CFR 50.91(a), the SNC analysis of the issue of no significant hazards consideration is presented below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed relocation of program details from TS 5.5.16 of each plant to their respective UFSARs is appropriate since these requirements do not satisfy the categories and criteria of 10CFR50.36(c), which specifies what items qualify for inclusion in the TS. The augmented inspections, which are performed in addition to required ASME Code Section Xl inspections or examinations, will continue to be performed as required by the UFSAR. The requirement to have an augmented inspection program remains in the TS. The plant systems and components to which the augmented inspections apply will not be operated in a different manner. The proposed relocation of the augmented inspection details does not involve a physical change to the plant or a change to the manner in which the plant is operated or controlled.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed changes do not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. As such, no new or different types of equipment will be installed, and the basic operation of installed plant systems and components, to which the augmented inspections apply, is unchanged. Hence, the proposed changes do not introduce any new accident initiators, nor do these changes reduce or adversely affect the capabilities of any plant structure or system in the performance of their safety function.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

Evaluation of Proposed Change E-9

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The proposed changes do not alter the manner in which safety limits or limiting safety system settings are determined. The safety analysis acceptance criteria are not affected by these proposed changes. Further, the proposed changes do not change the design function of any equipment assumed to operate in the event of an accident.

The proposed changes will not reduce a margin of safety because the relocation of details of the augmented inspection program to the plants respective UFSARs has no impact on any safety analysis assumptions, as indicated by the fact that the requirements do not meet the 10CFR50.36(c) criteria for inclusion in the TS. In addition, the details of the augmented inspections will be moved to the UFSARs and will continue to be performed as required by the UFSARs.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

4.4 Conclusion In conclusion, based on the considerations discussed above, SNC concludes: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

The proposed change would revise a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, and would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

6.0 REFERENCES

1. NUREG 75/034 (aka NUREG 0117), Supplement No. 1 to the Safety Evaluation Report by the Office of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Commission in the Matter of Alabama Power Company, Joseph M. Farley Nuclear Power Plant Units 1 and 2.

2. NUREG 1137, Safety Evaluation Report Related to the Operation of Vogtle Electric Generating Plant, Units 1 and 2, dated June 1985

3. Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, 58 FR 39132, dated July 22, 1993

Southern Nuclear Operating Company Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Relocate Augmented Piping Inspection Program Details from Technical Specifications to a Licensee Controlled Document a

FNP Proposed Technical Specification Changes (Marked-up Pages)

Programs and Manuals 5.5 Farley Units 1 and 2 5.5-13 Amendment No. 203 (Unit 1)

Amendment No. 199 (Unit 2) 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)

b.

Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;

c.

Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

d.

Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a.

A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

b.

A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

c.

A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.16 Main Steamline Inspection Program The three main steamlines from the rigid anchor points of the containment penetrations downstream to and including the main steam header shall be inspected. The extent of the inservice examinations completed during each inspection interval (IWA 2400, ASME Code, 1974 Edition,Section XI) shall provide 100 percent volumetric examination of circumferential and longitudinal pipe welds to the extent practical. The areas subject to examination are those defined in accordance with examination category C-G for Class 2 piping welds in Table IWC-2520.

In accordance with the augmented inservice inspection program for high energy lines outside of containment, examinations of welds in the main steam lines of each unit shall be performed to provide assurance of the continued

Programs and Manuals 5.5 Farley Units 1 and 2 5.5-13 Amendment No. 203 (Unit 1)

Amendment No. 199 (Unit 2) integrity of the piping systems over their service lifetime. These requirements apply to welds in piping systems or portions of systems located outside of containment where protection from the consequences of postulated ruptures is not provided by a system of pipe whip restraints, jet impingement barriers, protective enclosures and/or other measures designed specifically to cope with such ruptures.

Southern Nuclear Operating Company Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Relocate Augmented Piping Inspection Program Details from Technical Specifications to a Licensee Controlled Document b

VEGP Proposed Technical Specification Changes (Marked-up Pages)

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.16 MS and FW Piping Inspection Program This program shall provide for the inspection of the four Main Steam and Feedwater lines from the containment penetration flued head outboard welds, up to the first five-way restraint. The extent of the inservice examinations completed during each inspection interval (ASME Code Section XI) shall provide 100%

volumetric examination of circumferential and longitudinal welds to the extent practical. This augmented inservice inspection is consistent with the requirements of NRC Branch Technical Position MEB 3-1, "Postulated Break and Leakage Locations in Fluid System Piping Outside Containment," November 1975 and Section 6.6 of the FSAR.

In accordance with the augmented inservice inspection program for high energy lines outside of containment, examinations of welds in the main steam and main feedwater lines of each unit shall be performed to provide assurance of the continued integrity of the piping systems over their service lifetime. These requirements apply to welds in piping systems or portions of systems located outside of containment where protection from the consequences of postulated ruptures is not provided by a system of pipe whip restraints, jet impingement barriers, protective enclosures and/or other measures designed specifically to cope with such ruptures.

5.5.17 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI)

Topical Report (TR) NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, as modified by the following exceptions:

1.

Leakage rate testing for containment purge valves with resilient seals is performed once per 18 months in accordance with LCO 3.6.3, SR 3.6.3.6 and SR 3.0.2.

2.

Containment personnel air lock door seals will be tested prior to reestablishing containment integrity when the air lock has been used for containment entry. When containment integrity is required and the air lock has been used for containment entry, door seals will be tested at least once per 30 days during the period that containment entry(ies) is (are) being made.

3.

The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified Vogtle Units 1 and 2 5.5-16 Amendment No. 197 (Unit 1)

Amendment No. 180 (Unit 2)

Southern Nuclear Operating Company Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Relocate Augmented Piping Inspection Program Details from Technical Specifications to a Licensee Controlled Document a

FNP Revised Technical Specification Pages

Programs and Manuals 5.5 Farley Units 1 and 2 5.5-13 Amendment No. (Unit 1)

Amendment No. (Unit 2) 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)

b.

Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;

c.

Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

d.

Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a.

A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

b.

A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

c.

A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.16 Main Steamline Inspection Program In accordance with the augmented inservice inspection program for high energy lines outside of containment, examinations of welds in the main steam lines of each unit shall be performed to provide assurance of the continued integrity of the piping systems over their service lifetime. These requirements apply to welds in piping systems or portions of systems located outside of containment where protection from the consequences of postulated ruptures is not provided by a system of pipe whip restraints, jet impingement barriers, protective enclosures and/or other measures designed specifically to cope with such ruptures.

Southern Nuclear Operating Company Joseph M. Farley Nuclear Plant - Units 1 and 2 Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Relocate Augmented Piping Inspection Program Details from Technical Specifications to a Licensee Controlled Document b

VEGP Revised Technical Specification Pages

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.16 MS and FW Piping Inspection Program In accordance with the augmented inservice inspection program for high energy lines outside of containment, examinations of welds in the main steam and main feedwater lines of each unit shall be performed to provide assurance of the continued integrity of the piping systems over their service lifetime. These requirements apply to welds in piping systems or portions of systems located outside of containment where protection from the consequences of postulated ruptures is not provided by a system of pipe whip restraints, jet impingement barriers, protective enclosures and/or other measures designed specifically to cope with such ruptures.

5.5.17 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI)

Topical Report (TR) NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, as modified by the following exceptions:

1.

Leakage rate testing for containment purge valves with resilient seals is performed once per 18 months in accordance with LCO 3.6.3, SR 3.6.3.6 and SR 3.0.2.

2.

Containment personnel air lock door seals will be tested prior to reestablishing containment integrity when the air lock has been used for containment entry. When containment integrity is required and the air lock has been used for containment entry, door seals will be tested at least once per 30 days during the period that containment entry(ies) is (are) being made.

3.

The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief or alternative has been authorized by the NRC. At the discretion of the licensee, the containment concrete visual examinations may be performed during either power operation, e.g., performed concurrently with other containment inspection-related activities such as tendon testing, or during a maintenance/refueling outage.

(continued)

Vogtle Units 1 and 2 5.5-16 Amendment No. (Unit 1)

Amendment No. (Unit 2)