ML12097A248

License Amendment Request for a One-Time, 15-Month Extension to the Integrated Leak Rate Test Interval License Amendment Request No. 2012-03

ML12097A248
Person / Time
Site:	Oconee
Issue date:	04/03/2012
From:	Gillespie T Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR-12-03
Download: ML12097A248 (37)

Text

Duke T.PRESTON GILLESPIE, JR.

Vice President EEnergy Oconee Nuclear Station Duke Energy ON01 VP / 7800 Rochester Hwy.

Seneca, SC 29672 10 CFR 50.90 864-873-4478 864-873-4208 fax April 3, 2012 T.Gillespie@duke-energy.com U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Duke Energy Carolinas, LLC Oconee Nuclear Station, Unit 1 Renewed Facility Operating License Number DPR-38 Docket Number 50-269 License Amendment Request for a One-Time, 15-Month Extension to the Integrated Leak Rate Test Interval License Amendment Request No. 2012-03 In accordance with 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy) requests Nuclear Regulatory Commission (NRC) review and approval to amend the Technical Specifications (TS) of Renewed Facility Operating License No. DPR-38. The proposed change would allow for a one-time extension to the ten-year frequency of the Oconee Unit 1 containment leakage rate test (i.e., Integrated Leak Rate Test (ILRT) or Type A test). This test is required by Technical Specification (TS) 5.5.2 "Containment Leakage Rate Testing Program."

The proposed change would permit the existing ILRT frequency to be extended from ten years to approximately 11.25 years.

The proposed revision would avoid the necessity of performing a Type A test 13 months prior to the 10th year anniversary of the completion of the last Type A test (December 8, 2003). If granted, this revision would extend the period from 120 months (10 years) to no longer than 135 months between successive tests. In terms of refueling outages, this extension would move the performance of the next ILRT from the scheduled fall 2012 refueling outage (1EOC27) to the fall 2014 refueling outage (1EOC28).

The last Oconee Unit 1 ILRT was completed on December 8, 2003. The next ILRT is required by TS 5.5.2 to be performed no later than December 8, 2013, which is approximately 12 months after the conclusion of Oconee Unit 1 outage 1 EOC27. The proposed change would encompass the currently scheduled completion of 1EOC28, approximately 12 months beyond the present frequency. This request is for a 15 month extension, which bounds the time to begin the next operating cycle. This additional time is requested to allow flexibility in the schedule to address any potential extended down powers or forced outages or unforeseen issues that may arise during an outage without having to revise this request.

Regulatory evaluation (including the significant hazards consideration) and environmental considerations are provided in the Enclosure. The proposed change does not include any new commitments.

www. duke-energy.com

US Nuclear Regulatory Commission April 3, 2012 Page 2 In accordance with Duke Energy administrative procedures that implement the Quality Assurance Program Topical Report, these proposed changes have been reviewed and approved by the Plant Operations Review Committee. A copy of this LAR is being sent to the State of South Carolina in accordance with 10 CFR 50.91 requirements.

Duke Energy requests approval of this amendment request by September 15, 2012. Once approved, the amendment will be implemented within 60 days. Duke Energy will update applicable sections of the ONS Updated Final Safety Analysis Report (UFSAR), as necessary, and submit these changes in accordance with 10 CFR 50.71(e). There are no new commitments being made as a result of this proposed change. Inquiries on this proposed amendment request should be directed to Sandra Severance of the Oconee Regulatory Compliance Group at (864) 873-3466.

I declare under penalty of perjury that the foregoing is true and correct. Executed on April 3, 2012.

Sincerely, T. Preston Gillespie, Jr., Vice President, Oconee Nuclear Station

Enclosure:

1. Evaluation of Proposed Change Attachments:

1. Technical Specifications - Mark Ups

2. Technical Specifications - Reprinted Pages

US Nuclear Regulatory Commission April 3, 2012 Page 3 cc w/enclosure and attachments:

Mr. Victor McCree Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Mr. John Stang Senior Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8 G9A Washington, DC 20555 Mr. Andy Sabisch NRC Senior Resident Inspector Oconee Nuclear Station Susan E. Jenkins, Manager, Radioactive & Infectious Waste Management, SC Department of Health & Environmental Control 2600 Bull Street Columbia, SC 29201

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-1 ENCLOSURE I EVALUATION OF PROPOSED CHANGES

Subject:

Proposed License Amendment Request to Request a One-Time, 15-Month Extension to the Oconee Unit 1 Integrated Leak Rate Test Interval

1.

SUMMARY

DESCRIPTION

2. BACKGROUND

3. DETAILED DESCRIPTION OF PROPOSED CHANGES

4. TECHNICAL EVALUATION

5. REGULATORY EVALUATION

• Significant Hazards Consideration

• Applicable Regulatory Requirements/Criteria

• Precedent

• Conclusion

6. ENVIRONMENTAL CONSIDERATION

7. REFERENCES

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-2

1.

SUMMARY

DESCRIPTION Duke Energy Carolinas, LLC (Duke Energy) requests to amend the Renewed Facility Operating License DPR-38 for the Oconee Nuclear Station, Unit 1 (ONS 1) to revise Technical Specifications (TS), 5.5.2 "Containment Leakage Rate Testing Program,"

requirements. The proposed change would allow for a one-time extension to the ten-year frequency of the Oconee Unit 1 containment leakage rate test (i.e., Integrated Leak Rate Test (ILRT) or Type A test). This test is required by Technical Specification (TS) 5.5.2 "Containment Leakage Rate Testing Program." The proposed change would permit the existing ILRT frequency to be extended from ten years to approximately 11.25 years.

The proposed revision would avoid the necessity of performing a Type A test 13 months prior to the 10th year anniversary of the completion of the last Type A test (December 8, 2003). If granted, this revision would extend the period from 120 months (10 years) to no longer than 135 months between successive tests. In terms of refueling outages, this extension would move the performance of the next ILRT from the scheduled fall 2012 refueling outage (1 EOC27) to the fall 2014 refueling outage (1 EOC28).

Duke Energy is proposing this revision based on the satisfactory containment leakage rate history and containment visual examination history at ONS 1. Additional insights from a risk analysis of this extension support the deterministic argument of extending the inspection interval 15 months. 1EOC28 is currently scheduled to end approximately twelve months after the current ILRT due date. This request for a 15-month extension would bound the time to reach 1 EOC28 and provide additional time to allow flexibility in the schedule to address any potential extended down powers, forced outages or unforeseen issues that may arise during that outage and the intervening time before 1 EOC28 without having to revise this request. Including the ILRT in 1 EOC27, which is scheduled for October 2012, would result in its performance approximately 13 months prior to the ILRT ten year anniversary due date.

2. BACKGROUND 2.1 Description of Primary Containment System The primary containment is described in Updated Final Safety Analysis Report (UFSAR)

Sections 1.2.2.3, 3.8.1, and 6.2.1.1.1.

The containment is a reinforced concrete structure with a cylindrical wall, a flat foundation mat, and a shallow dome roof. The containment design includes un-grouted tendons where the cylinder wall is pre-stressed with a post tensioning system in the vertical and horizontal directions, and the dome roof is pre-stressed using a three-way tensioning system. The inside surface of the containment is lined with a carbon steel liner to ensure a high degree of leak tightness during operating and accident conditions.

The structure consists of a post-tensioned reinforced concrete cylinder and dome connected to and supported by a massive reinforced concrete foundation slab. The entire interior surface of the structure is lined with a welded ASTM A36 steel plate to

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-3 assure a high degree of leak tightness. Numerous mechanical and electrical systems penetrate the Reactor Building wall through welded steel penetrations. The mechanical penetrations and access openings are designed, fabricated, inspected, and installed in accordance with Subsection B,Section III, of the ASME Pressure Vessel Code. All piping and ventilation penetrations are of the rigid welded type and are solidly anchored to the Reactor Building wall or foundation slab, thus precluding any requirements for expansion bellows.

Principal dimensions are as follows:

• Inside Diameter 116 ft.

• Inside Height (Including Dome) 208-1/2 ft.

• Vertical Wall Thickness 3-3/4 ft.

• Dome Thickness 3-1/4 ft.

• Foundation Slab Thickness 8-1/2 ft.

• Liner Plate Thickness 1/4 in.

• Internal Free Volume 1,836,000 cu ft. (as-built) 2.2 Testingq Requirements of 10 CFR 50, Appendix J The testing requirements of 10 CFR 50, Appendix J, provide assurance that leakage from the containment, including systems and components that penetrate the containment, does not exceed the allowable leakage values specified in the TS. 10 CFR 50, Appendix J also ensures that periodic surveillance of reactor containment penetrations and isolation valves is performed so that proper maintenance and repairs are made during the service life of the containment and the systems and components penetrating primary containment. The limitation on containment leakage provides assurance that the containment would perform its design function following an accident up to and including the plant design basis accident. Appendix J identifies three types of required tests: (1) Type A tests, intended to measure the primary containment overall integrated leakage rate; (2) Type B tests, intended to detect local leaks and to measure leakage across pressure-containing or leakage limiting boundaries (other than valves) for primary containment penetrations; and (3) Type C tests, intended to measure containment isolation valve leakage rates. Type B and C tests identify the vast majority of potential containment leakage paths. Type A tests identify the overall (integrated) containment leakage rate and serve to ensure continued leakage integrity of the containment structure by evaluating those structural parts of the containment not covered by Type B and C testing.

In 1995, 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," was amended to provide a performance-based Option B for the containment leakage testing requirements. Option B requires that test intervals for Type A, Type B, and Type C testing be determined by using a performance-based approach. Performance-based test intervals are based on consideration of the operating history of the component and resulting risk from its failure. The use of the term "performance-based" in 10 CFR 50 Appendix J refers to both the performance history necessary to extend test intervals as well as to the criteria necessary to meet the requirements of Option B.

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-4 Also in 1995, Regulatory Guide (RG) 1.163 was issued. The RG endorsed Nuclear Energy Institute (NEI) 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J" with certain modifications and additions. Option B, in concert with RG 1.163 and NEI 94-01, Revision 0, allows licensees with a satisfactory ILRT performance history (i.e., two consecutive, successful Type A tests) to reduce the test frequency for the containment Type A (ILRT) test from three tests in 10 years to one test in 10 years. This relaxation was based on an NRC risk assessment contained in NUREG-1493, "Performance-Based Containment Leak-Test Program," and Electric Power Research Institute (EPRI) TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals," both of which showed that the risk increase associated with extending the ILRT surveillance interval was very small. In addition to the 10-year ILRT interval, provisions for extending the test interval an additional 15 months was considered in the establishment of the intervals allowed by RG 1.163 and NEI 94-01, but that this "should be used only in cases where refueling schedules have been changed to accommodate other factors."

2.3 Current ONS Technical Specification Requirements On October 30, 1996, the NRC issued Amendment No. 218 to Facility Operating License No. DPR-38 for ONS 1. This amendment was in response to the application dated August 12, 1996, and supplement dated September 10, 1996, ML012050049. The amendment revised the TS associated with the containment leak-rate tests by implementing 10 CFR Part 50, Appendix J, Option B for Type A leak-rate tests only. The performance of Type B and C leak-rate tests remained under Option A of 10 CFR Part 50, Appendix J.

On July 28, 2011, the NRC issued Amendment No. 375 to Renewed Facility Operating License No. DPR-38 for ONS 1. The amendment was in response to the application dated July 14, 2010, ML11186A906. This amendment revised the TS to adopt technical specification task force technical change Traveler 52, Revision 3, to implement 10 CFR Part 50, Appendix J, Option B for Type B and C leak-rate tests.

3. DETAILED DESCRIPTION OF PROPOSED CHANGES 3.1' Current Technical Specification Requirement ONS TS 5.5.2, "Containment Leakage Rate Testing Program," currently states:

A program shall establish the leakage rate testing of the containment as requiredby 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordancewith the guidelines contained in Regulatory Guide 1.163, "Performance-BasedContainmentLeak-Test Program," dated September 1995. Containment system visual examinations requiredby Regulatory Guide 1.163, Regulatory Position C.3, shall be performed as follows:

1. Accessible concrete surfaces and post-tensioning system component surfaces of the concrete containment shall be visually examined priorto initiating SR 3.6.1.1 Type A

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-5 test. These visual examinations, or any portion thereof, shall be performed no earlier than 90 days priorto the start of refueling outages in which Type A tests will be performed. The validity of these visual examinations will be evaluated should any event or condition capable of affecting the integrity of the containment system occur between the completion of the visual examinations and the Type A test.

2. Accessible interiorand exteriorsurfaces of metallic pressure retainingcomponents of the containmentsystem shall be visually examined at least three times every ten years, including during each shutdown for SR 3.6.1.1 Type A test, prior to initiating the Type A test.

The calculatedpeak containment internalpressure for the design basis loss of coolant accident, Pa, is 59 psig. The containment design pressure is 59 psig.

The maximum allowable containment leakage rate, La, at Pa, shall be 0.20% of the containmentair weight per day.

Leakage rate acceptance criterionis:

a. Containmentleakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteriaare 5 0.60 La for the Type B and C tests, and < 0.75 La for Type A tests; The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

Nothing in these Technical Specifications shall be construed to modify the testing Frequenciesrequired by 10 CFR 50, Appendix J.

3.2 The Proposed Change The proposed change would revise the initial paragraph of TS 5.5.2 by the addition of a specific requirement for Unit 1 while maintaining the current requirement for Units 2 and 3 (changes underlined), as shown below:

A program shall establish the leakage rate testing of the containment as requiredby 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The next Unit 1 ILRT following the December 8, 2003 test shall be performed no later than March 8, 2015. This program shall be in accordancewith the guidelines contained in Regulatory Guide 1.163, "Performance-BasedContainment Leak-Test Program," dated September 1995. Containment system visual examinations requiredby Regulatory Guide 1.163, Regulatory Position C.3, shall be performed as follows:

1. Accessible concrete surfaces and post-tensioning system component surfaces of the concrete containment shall be visually examined priorto initiatingSR 3.6.1.1 Type A test. These visual examinations,or any portion thereof, shall be performed no earlier

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-6 than 90 days priorto the start of refueling outages in which Type A tests will be performed. The validity of these visual examinations will be evaluated should any event or condition capable of affecting the integrity of the containment system occur between the completion of the visual examinations and the Type A test.

2. Accessible interiorand exteriorsurfaces of metallic pressure retainingcomponents of the containmentsystem shall be visually examined at least three times every ten years, including during each shutdown for SR 3.6.1.1 Type A test, prior to initiating the Type A test.

The calculatedpeak containment internal pressure for the design basis loss of coolant accident, Pa, is 59 psig. The containment design pressure is 59 psig.

The maximum allowable containment leakage rate, La, at Pa, shall be 0. 20% of the containment airweight per day.

Leakage rate acceptance criterion is:

a. Containment leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordancewith this program, the leakage rate acceptance criteriaare < 0.60 La for the Type Band C tests, and < 0 75 La for Type A tests; The provisions of SR 3.0.3 are applicable to the ContainmentLeakage Rate Testing Program.

Nothing in these Technical Specifications shall be construed to modify the testing Frequenciesrequired by 10 CFR 50, Appendix J. contains existing TS page 5.0-7 marked up to show the proposed changes to TS 5.5.2. Attachment 2 contains the re-type of the proposed TS changes.

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-7

4. TECHNICAL EVALUATION In addition to the periodic integrated leakage rate testing of the Reactor Containment, various other inspections and tests are performed on an ongoing basis to help assure primary containment integrity. The relevant basis, history, and results of these inspections and tests, as well as discussion of other, generic areas of interest, are included in the subsections below to aid in the review of this request. These include

• 4.1 - Second Interval Containment Inspection Plan

• 4.2 -Augmented Inspections

• 4.3 - Recent IWE/IWL Inspection Results

• 4.4 - Loss of Tendon Prestress

• 4.5 - Inaccessible Areas

• 4.6 - Containment Coatings

• 4.7 - Previous ILRT Results

• 4.8 - Appendix J Type B and Tpe C Performance Testing

• 4.9 - IN 92-20, Inadequate Local Leak Rate Testing

• 4.10- Supplemental Inspections

• 4.11 - Plant Specific Confirmatory Analysis.

As shown below, the combined results of these tests and inspections provide a high degree of assurance of continued primary containment integrity.

4.1 Second Interval Containment Inservice Inspection Plan The Oconee UFSAR classifies the Reactor Building as a Class 1 structure. Class 1 structures are those which prevent uncontrolled release of radioactivity and are designed to withstand all loadings without loss of function. The current American Society of Mechanical Engineers (ASME) Code component classifications did not exist at the time of plant design and construction. The ASME Class CC and MC designations were added in later editions of the Code. For the purposes of this plan, Class CC components will consist of the concrete Reactor Building and the Reactor Building Un-bonded Post-Tensioning System, and Class MC components will consist of the metallic shell and penetration liners.

Inspection Interval and Inspection Periods The Second/Third Containment In-service Inspection Intervals for ONS 1 are shown below. Please note that these intervals do not coincide with In-service Inspection Intervals for ASME Class 1, 2, and 3 systems and components. The term "EOC" used below is an abbreviation for "End of Cycle" and the associated number indicates the sequential refueling outage following initial operation of the unit. These inspection intervals and periods have been modified in accordance with the provisions of Duke Energy Corporation Request for Alternative Serial #03-GO-010.

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-8 IWE - Metal Containments and Liners of Concrete Containments In accordance with TS 5.5.2, accessible interior and exterior surfaces of metallic pressure retaining components of the containment system shall be visually examined at least three times every ten years, including during each shutdown for TS Surveillance Requirement (SR) 3.6.1.1 Type A test, prior to initiating the Type A test.

The following table documents the Inspection Intervals and periods to meet the requirements of TS 5.5.2 and ASME Section Xl, Table IWE-2500.

Second/Third Containment In-service Inspection Interval Unit I (IWE)

(See Notes)

Interval 2 Start Date End Date 07/15/2005 07/15/2008 07/15/2011. 07/14/2014 1st Period 2nd Period 3rd Period Outage 1 (EOC 23) Outage 3 (EOC 25) Outage 5 (EOC 27)

Note 4 Outage 2 (EOC 24) Outage 4 (EOC 26)

Interval 3*

Start Date End Date 07/15/2014 07/15/2017 07/15/2020 07/14/2024 1st Period 2nd Period 3rd Period Outage 1 (EOC 28) Outage 3 (EOC 30) Outage 4 (EOC 31)

Outage 2 (EOC 29) Outage 5 (EOC 32)

• The scheduled dates for the Third Containment In-service Inspection Intervals are proposed dates in that the Third Interval inspection program has not been approved at this time.

If the IWE-2500, Table IWE-2500-1, Category E-A, Item E1.11 examination is to be credited towards satisfying the examinations required by 10 CFR 50 Appendix J, the examination shall be performed during the refueling outage during which a Type A test is to be performed, prior to the start of the Type A test. Duke Energy intends to credit ASME Section Xl, Table IWE-2500-1, Item E1.11 visual exams towards satisfying the requirements of 10 CFR 50 Appendix J.

IWL - Concrete Components And Post-Tensioning Systems In accordance with TS 5.5.2, accessible concrete surfaces and post-tensioning system component surfaces of the concrete containment shall be visually examined prior to initiating SR 3.6.1.1 Type A test. These visual examinations, or any portion thereof, shall

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-9 be performed no earlier than 90 days prior to the start of refueling outages in which Type A tests will be performed. The validity of these visual examinations will be evaluated should any event or condition capable of affecting the integrity of the containment system occur between the completion of the visual examinations and the Type A test.

When possible, Duke intends to credit the IWL concrete examinations towards meeting the above TS requirement. IWL concrete and post-tensioning system examinations and tests are also performed in accordance with the schedule provided in the table below.

Unit 1 (IWL)

(See Notes)

Interval 2 IWL Period 1: 8/4/05 - 8/4/07 (35th Year Exams)

IWL Period 2: 8/4/10 - 8/4/12 (40th Year Exams)

Interval 3 IWL Period 1: 8/4/15 - 8/4/17 (45th Year Exams)

IWL Period 2: 8/4/20 - 8/4/22 (50th Year Exams)

Notes:

1. IWL Periods for Unit 1 are based on a repeating 5 year schedule (+/- 12 month) following the completion of the containment Structural Integrity Test, as required by IWL-2410 and IWL-2420. The Unit 1 Structural Integrity Test was completed on 8/4/1971. Initial post-tensioning operations were completed in November, 1970.

2. IWE Examinations are scheduled for the Second Inspection Interval in accordance with ASME Section XI Inspection Plan B, Table IWE-2412-1, as modified in accordance with the provisions of Duke Energy Corporation Relief Request Serial #03-GO-010.

3. ISI Interval 2 has been reduced by 12 months, as permitted by IWA -2430(d)(1).

4. IWE Period 2 has been reduced by 12 months, as permitted by IWA-2430(d)(3).

4.2 Containment Surfaces Subiect To Augqmented Inspections In accordance with the ONS 1 "Second Interval Containment In-service Inspection Plan,"

the following augmented inspections are currently required to be performed each Inspection Period. As noted in the table, the most recent examination results have been determined to be acceptable in all cases. Following the Augmented Inspection Table is a discussion of those examinations that required evaluation in accordance with the requirements of Section Xl, IWE and IWL and the results. All other examination results (including augmented examinations) were acceptable by examination.

Enclosure 1 - Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page El-1O Please note that the scope of augmented examinations for the 3rd Containment ISI Interval have not yet been determined and may vary from those listed below. The scope of augmented examinations (IWE-2500, Table IWE-2500-1, Examination Category E-C) for the 3rd Inspection Interval shall be determined in accordance with requirements of the applicable Section XI Code of Record to be used during the 3rd Containment ISI Interval.

Second Interval Containment In-service Inspection Plan Augmented Inspections Item Component Component Inspection/ Comments/Miscellaneous Examination Number ID Description NDE Code Requirements Results Required E04.11.0001 1-SCV-001 Visible VT-1 (Metal Examination is limited to Acceptable.

Surfaces Containment) surfaces at embedment zones within 2" of the basement floor at elevation 777+6 (Nom.) at accessible locations shown on inspection drawings. Added as a result of PIP's 0-096-2414 and 1-099-2317. See Note 2.

E04.11.0002 1-MOBR- Moisture VT-1 (Metal Added as a result of PIP's Acceptable.

001 Barrier Containment) 0-096-2414 and 1-099 2317.

May be performed in conjunction with Item E1.30 examination each Period.

See Note 2.

E04.11.0003 1-MOBR- Moisture VT-1 (Metal Added as a result of PIP's Acceptable.

005 Barrier Containment) 0-096-2414 and 1-099-2317.

May be performed in conjunction with Item E1.30 examination each Period.

See Note 2.

E04.11.0004 1-MOBR- Moisture VT-1 (Metal Added as a result of PIP's Acceptable.

010 Barrier Containment) 0-096-2414 and 1-099-2317.

May be performed in conjunction with Item E1.30 examination each Period.

See Note 2.

Enclosure 1 - Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page El-1I Item Component Component Inspection/ Comments/Miscellaneous Examination Number ID Description NDE Code Requirements Results Required E04.11.0005 1-SCV-011 Visible VT-1 (Metal Examination is limited to Acceptable.

Surfaces Containment) surfaces of the containment liner plate within inspection port located at azimuth 1350 on the basement floor at elevation 777'+6." Note:

Inspection port plug must be removed to permit visual examination (See drawings

1. 0-67A & 0-67A-005).

Added as a result of PIP 0-96-2414. May be performed in conjunction with Item E1.11 examination. See Note 2.

E4.12.0001 1-GRID-001 Surface UT Note: Inspection port plug Acceptable.

Area must be removed to permit Grid examination (See drawings

1. O-67A & 0-67A-005).

Added as a result of PIP 0-96-2414. See Notes 2 and 6.

E4.12.0002 1-GRID- Surface UT Added as a result of PIP Acceptable.

A159 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0003 1-GRID- Surface UT Added as a result of PIP Acceptable.

A160 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0004 1-GRID- Surface UT Added as a result of PIP Acceptable.

A161 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0005 1-GRID- Surface UT Added as a result of PIP Acceptable.

A162 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0006 1-GRID- Surface UT Added as a result of PIP Acceptable.

A163 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0007 1-GRID- Surface UT Added as a result of PIP Acceptable.

A164 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0008 1-GRID- Surface UT Added as a result of PIP Acceptable.

A165 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0009 1-GRID- Surface UT Added as a result of PIP Acceptable.

A166 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0010 1-GRID- Surface UT Added as a result of PIP Acceptable.

A167 Area 0-08-01395. See Notes 2 Grid and 6.

Enclosure 1 - Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page El-12 Item Component Component Inspection/ Comments/Miscellaneous Examination Number ID Description NDE Code Requirements Results Required E4.12.0011 1-GRID- Surface UT Added as a result of PIP Acceptable.

A168 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0012 1-GRID- Surface UT Added as a result of PIP Acceptable.

A204 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0013 1-GRID- Surface UT Added as a result of PIP Acceptable.

A205 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0014 1-GRID- Surface UT Added as a result of PIP Acceptable.

A206 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0015 1-GRID- Surface UT Added as a result of PIP Acceptable.

A207 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0016 1-GRID- Surface UT Added as a result of PIP Acceptable.

A208 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0017 1-GRID- Surface UT Added as a result of PIP Acceptable.

A210 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0018 1-GRID- Surface UT Added as a result of PIP Acceptable.

A21 1 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0019 1-GRID- Surface UT Added as a result of PIP Acceptable.

A212 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0020 1-GRID- Surface UT Added as a result of PIP Acceptable.

A213 Area 0-08-01395. See Notes 2 Grid and 6.

E4.12.0021 1-GRID- Surface UT Added as a result of PIP Acceptable.

A214 Area 0-08-01395. See Notes 2 Grid and 6.

The following notes from the Second Interval Containment Inservice Inspection Plan and Problem Investigation Program (PIP) documents are referenced in the Comments column in the table above and are included here for additional information.

Note 2. Exam may be discontinued after 2 consecutive periods if the requirement of IWE-2420(c) has been met.

Note 6. Accessible surfaces of the containment metallic liner plate shall be examined and the location of the minimum wall thickness shall be located using the coordinate system shown on applicable IS[ drawing. Subsequent examinations need only be performed at the identified minimum wall thickness location.

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-13 PIP Report #0-99-2317, Degradation of moisture barrier:

This report documents adverse conditions observed on moisture barriers at the containment liner plate embedment zone. A summary of the observed conditions, evaluations conducted, and corrective actions is documented in a letter to the NRC, dated October 13, 1999 [submitted pursuant to 10 CFR 50.55a(b)(2)(x)(A) - now 10 CFR 50.55a(b)(2)(ix)(A)].

PIP Report #0-96-2414, Majority of sealant along basement slab/liner plate interface and between edge of slab and columns, walls, and foundations is missing or degraded:

This report documents the results of inspections performed by site engineering during which degradation of moisture barriers (sealant) was observed at the interface between the containment liner plate and the interior concrete base slab.

Sealant degradation was also noted at other interior joints in the base slab concrete. As a result of these observed conditions, moisture barriers were corrected and permanent (removable) inspection ports were added in the base concrete slab at the containment liner plate interface to allow for continued monitoring of conditions immediately behind the liner plate/base slab concrete interface. Subsequent examination of these areas has been performed in accordance with ASME Code, Section Xl, IWE-2500, Table IWE-2500-1, Examination Category E-C, and the results of these examinations performed during the Second Containment ISI Interval have been acceptable.

PIP Report #0-08-01395, Documentation of Containment Integrity Assessment results from PIP G-06-00465:

During Steam Generator Replacement Project activities, hydrolazing was used to remove concrete from the containment at the location of the'temporary construction opening. During this activity, a significant amount of water was introduced into the space between the liner plate and the interior surface of the concrete, beneath the temporary opening. This additional water could increase the risk of potential corrosion of the liner plate in these areas where any gaps exist between the concrete and the containment liner plate. To address this concern, the Oconee Containment ISI Plan was revised to add ultrasonic thickness measurement of liner plate areas beneath the repaired opening in accordance with the ASME Code,Section XI, Category E-C, Item E4.12.

Locations of wall thickness examinations are identified in the Oconee Containment ISI Plan. Results of these examinations performed during the Second Containment ISI Interval have revealed no detectable wall thickness loss.

4.3 Recent IWE/IWL Inspection Results (Since start of Inspection Interval 2 on July 15, 2005)

The following is a summary of results of examinations that required evaluation in accordance with the requirements of Section Xl, IWE and IWL. All other examination results (including augmented examinations) were acceptable by examination.

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-14 IWL Examination Results:

During the 35th year post-tensioning system surveillance, the following conditions were observed:

1. Tendon Mark #62H21 was found to contain one gallon of free water. The water was tested and found to be slightly basic. No corrosion was found on anchorage hardware. The percent of water by lab analysis was less than 10%. Based on an evaluation of the grease results as tested, all samples were within limits for all criteria. This condition has been identified previously, as documented in a letter to the NRC, dated April 9, 2004 [submitted pursuant to 10 CFR 50.55a(b)(2)(viii)(D)(1)].

The condition of tendon #62H21 was evaluated as acceptable.

2. The absolute difference between the amount of corrosion protection medium removed and the amount replaced exceeded 10% of the tendon net duct volume for the following tendons. Internal Operating Experience has demonstrated that loss of sheathing filler grease, by itself, does not indicate that tendon degradation has occurred, and actions have been taken to address the generic issue of grease leakage from tendon sheaths. Examination results indicated that tendon metallic components remained adequately protected by grease coverage, even for those tendons where the tendon sheath is not completely filled with grease. All other examination and test results for the tendons identified below were acceptable.

% Difference Between Volume Installed Tendon Mark No. vs. Removed (Net Duct Volume) 12V20 10.6 23V23 12.5 62H23 18.5 46H66 11.3 2D37 18.3 1D31 13.0 1D03 20.3 2D31 22.8 1D07 42.6*

3D07 18.1

• Tendon 1 D07 is a Dome tendon. Although the amount of grease leakage was higher for this tendon (compared to other tested tendons), the metallic components were visually examined and were found to be acceptable. The tendon sheath was refilled with grease following examination and testing.

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-15 IWE Examination Results:

During refueling outage 1EOC25 (Interval 2, Period 2), the following conditions were observed:

PIP Report #0-09-08102 documented a lack of full thread engagement on electrical penetration EE10 bolting reported during ISI Containment Inspection (Refueling Outage 1 EOC25)

When performing the Bolted Connection VT-1 visual examination of Item

1. E08.10.0067 (electrical penetration Mk. EEl0) the inspector documented the results as unacceptable because 5 (of 12) bolts did not have full thread engagement.

Engineering determined that, due to the configuration of this penetration, the bolts in question have no design tension loads under accident conditions other than initial tensioning due to torqueing. The bolts attach a flange to the sleeve which is in additional compression under accident pressure and under ILRT pressure. The penetration is capable of performing its design function in the current condition and no corrective action was recommended because of concerns that any adjustment of the bolting could affect the leak-tightness of the metallic o-rings.

4.4 Loss of Tendon Prestress PIP G-06-00465 documents the results of the Containment Integrity Assessment, which was approved on March 13, 2008. Section 12.6, Loss of Tendon Prestress due to Wire Relaxation, Concrete Creep/Shrinkage, states that "Oconee operating experience has not detected tendon prestress loss in excess of prescribed limits since the initial implementation of the Containment ISI Program."

Tendon prestress losses will continue to be monitored through the performance of testing required by the ASME Code, Section Xl, Subsection IWL.

4.5 Inaccessible Areas For Class MC and CC applications, Duke Energy shall evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. For each inaccessible area identified, Oconee Nuclear Station shall provide the following in the ISI Summary Report, as required by 10 CFR 50.55a(b)(2)(viii)(E) and 10 CFR 50.55a(b)(2)(ix)(A):

• A description of the type and estimated extent of degradation, and the conditions that led to the degradation;

• An evaluation of each area, and the result of the evaluation, and;

• A description of necessary corrective actions.

Duke Energy has not needed to implement any new technologies to perform inspections of any inaccessible areas at this time. However, Duke Energy actively participates in

Enclosure 1 - Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-16 various nuclear utility owners groups, ASME Code committees, and with NEI to maintain cognizance of ongoing developments within the nuclear industry. Industry operating experience is also continuously reviewed to determine its applicability to ONS.

Adjustments to inspection plans and availability of new, commercially available technologies for the examination of the inaccessible areas of the containment would be explored and considered as part of these activities.

4.6 Containment Coatings Program The primary purpose of containment coatings, from an ILRT perspective, is to provide corrosion protection for the carbon steel liner plate to allow it to maintain its pressure retaining capability. The safety related coatings applied to the liner plate at Duke Energy nuclear stations are considered to be Service Level I, defined in Nuclear System Directive (NSD) 318, "Coating Program," as coatings applied to all exposed surface areas within the primary containment facilities which are required to withstand a Loss-Of-Coolant Accident (LOCA) environment.

Duke Energy has implemented controls for the procurement, application and maintenance of Service Level I protective coatings used inside containment in a manner that is consistent with the licensing basis and regulatory requirements applicable to Oconee.

The original liner plate coatings, consisting of a prime coat of inorganic zinc (IOZ) and a modified phenolic finish coat, were supplied by the Carboline Company and have been successfully tested by Carboline to withstand anticipated LOCA conditions. Carboline also supplies the Service Level I substitute coatings (epoxy mastic) now used for new applications and repair/replacement activities inside containment. The substitute coatings when used for maintenance over the original coatings were tested, with the appropriate documentation, to demonstrate a qualified coating system.

Condition assessments of Service Level I coatings used inside containment are performed during each refueling outage. If localized areas of degradation are identified, those areas are evaluated and scheduled for repair or replacement as necessary. The observed liner plate coating degradation at Oconee in the last ten years has typically consisted of the finish coat pulling away, or delaminating, from the IOZ primer. This delamination does not pose a liner plate corrosion protection issue because the remaining IOZ provides a sufficient corrosion barrier.

At the end of the last Unit 1 refueling outage (1 EOC26 - spring 2011), minimal degradation of the Unit 1 liner plate coating was noted. The 1EOC26 Coating Inspection Form, approved 06/01/2011, identified approximately 19 ft 2 of degraded finish coatings and approximately 5 ft2 of exposed zinc related to the liner plate. As noted above, exposed zinc continues to provide a sufficient corrosion barrier, and there is no visible corrosion present.

In summary, there are negligible amounts of liner plate coatings degradation in ONS 1, and the small amount that does exist poses minimal to no corrosion protection issues for the liner plate. As discussed above, condition assessments of ONS 1 containment coatings are performed every refueling outage and, based on ONS operating

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-17 experience, no coatings related containment structure leakage is expected to result from extending the next ILRT to 1EOC28 in fall 2014.

4.7 Previous ILRT Results Previous ILRT testing confirmed that the ONS 1 containment structure leakage is acceptable, with considerable margin, with respect to the TS acceptance criterion of 0.2% of containment air weight at the design basis loss of coolant accident pressure (La).

Since the last two ONS 1 Type A as-found results, as shown in the following table, were less than 1.0 La, a test frequency of at least once per 10 years would be in accordance with NEI 94-01, Revision 0.

The last ILRT was completed on December 8, 2003, after the installation of the replacement steam generators and closure of the construction opening made in the containment structure to support the replacement of the steam generators.

No modifications that require a Type A test are planned prior to 1 EOC28 when the next Type A test will be performed under this proposed change. Any unplanned modifications to the containment prior to the next scheduled Type A test would be subject to the special testing requirements of Section IV.A of 10 CFR 50, Appendix J. There have been no pressure or temperature excursions in the containment which could have adversely affected containment integrity. There is no anticipated addition or removal of plant hardware within containment which could affect leak-tightness that would not be challenged by local leak rate testing. Following the approval of this licensing amendment, the next ONS 1 ILRT must be performed on or before March 8, 2015.

ILRT Performance Results As-found results Allowable TS ONS 1 ILRT (wt %/day UCL) criterion Completion Date Note 1 (< 0.75 La) Test Pressure (psig) 12/8/03 0.0973 0.1875 60 1/17/93 0.1544 0.1875 60.57 5/26/90 0.1581 0.1875 59 4/7/86 0.1734 0.1875 59 8/3/83 0.152 0.1875 59 2/11/80 0.0222 0.0775 29.5 3/25/76 0.0422 0.0773 29.5 8/2/71 0.0475 0.25 59 (Initial ILRT) 0.0147 N/A 29.5 Note 1: The maximum allowable containment leakage rate, La, at Pa, shall be 0.20% of the containment air weight per day following the adoption of alternate source term in 2006. Prior to this the allowable containment leakage rate was 0.25% of the containment air weight per day.

Enclosure 1 - Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-18 4.8 Type B and C Testing Program The ONS 1 Appendix J, Type B and Type C testing program requires testing of electrical penetrations, airlocks, hatches, flanges, and valves within the scope of the program as required by 10 CFR 50, Appendix J, Option B and TS 5.5.2. The Type B and Type C testing program consists of local leak rate testing of penetrations with a resilient seal, double gasket man ways, hatches and flanges, and containment isolation valves that serve as a barrier to the release of the post-accident containment atmosphere.

On July 28, 2011, the NRC issued Amendment No. 375 to Renewed Facility Operating License DPR-38 for ONS 1. This amendment revised the TS to adopt technical specification task force technical change Traveler 52, Revision 3, to implement 10 CFR Part 50, Appendix J, Option B for Type B and C leak-rate tests. Prior to this amendment, the Type B and C testing program was conducted in accordance with 10 CFR 50 Appendix J, Option A. Under Option A, all penetrations were tested at the minimum frequency of 30 months.

A review of the Type B and Type C test results from the December 2003 through June 2011 and their comparison with the allowable leakage rate was performed. The currently established Type B and Type C leakage acceptance criterion is 212,402 standard cubic centimeters per minute (sccm). The minimum pathway leak rate summary totals for this time period shows minimum pathway leakage to be less than 11.72% of the limit as shown in the following table, Leak Rate History And Reactor Building Leak Rate Verification Results (PT/1/A/0150/034). It should be noted that all LLRT testing has been performed in accordance with Option A since the December 2003 ILRT.

As discussed in NUREG-1493, Type B and Type C tests can identify the vast majority (greater than 95%) of all potential containment leakage paths. This amendment request does not affect the scope, performance, or scheduling of Type B or Type C tests. Type B and Type C testing will continue to provide a high degree of assurance that containment integrity is maintained.

The fall 2012 outage will be the first ONS 1 refueling outage since the adoption of Option B for Type B and C tests on July 28, 2011. Transition from the prescriptive testing requirements of Option A to the performance-based requirements of Option B is in progress and will include the following: the establishment of extended test intervals for Type B and C tested components shall be performed in accordance with NEI 94-01 Revision 0 Sections 10.2.1.2 and 10.2.1.4 for Type B and Sections 10.2.3.2 and 10.2.3.4 for Type C tested components. As-found testing for Type B and Type C tested components shall also be performed for those components that will establish extended test intervals in accordance with NEI 94-01, Revision 0, Section 10.2.1.3 for Type B and Section 10.2.3.3 for Type C tested components. Containment airlocks shall be tested in accordance with NEI 94-01, Revision 0, Section 10.2.2.

- Evaluation of Proposed Changes License Amendment Request No. 201 2-03 April 3, 2012 Page El-19 LEAK RATE HISTORY AND REACTOR BUILDING LEAK RATE VERIFICATION RESULTS (PT/I IAIOI501034)

Acceptance Criteria Percentage of Date Result (scorn) (scorn) Acceptance Criteria 12/9/11 15,273 212,402 7.19%

10/14/11 16,573 212,402 7.80%

6/20/11 17,262 212,402 8.13%

6/6/11 16,138 212,402 7.60%

1/14/11 12,239 212,402 5.76%

1/13/11 12,239 212,402 5.76%

1/12/11 12,219 212,402 5.75%

1/10/11 12,219 212,402 5.75%

1/6/11 12,919 212,402 6.08%

11/9/10 12,919 212,402 6.08%

7/29/10 13,604 212,402 6.40%

5/24/10 12,864 212,402 6.06%

2/17/10 12,969 212,402 6.11%

11/28/09 18,409 212,402 8.67%

11/26/09 24,900 212,402 11.72%

10/5/09 23,929 212,402 11.27%

9/15/09 23,966 212,402 11.28%

7/14/09 19,266 212,402 9.07%

7/9/09 19,068 212,402 8.98%

4/21/09 19,070 212,402 8.98%

11/15/08 19,150 212,402 9.02%

9/29/08 21,348 212,402 10.05%

8/12/08 21,773 212,402 10.25%

6/9/08 21,988 212,402 10.35%

5/26/08 19,790 212,402 9.32%

4/15/08 12,780 212,402 6.02%

2/7/08 12,789 212,402 6.02%

1/2/08 13,404 212,402 6.31%

8/28/07 13,404 212,402 6.31%

7/26/07 14,389 212,402 6.77%

4/18/07 20,089 212,402 9.46%

4/11/07 19,929 212,402 9.38%

12/4/06 12,513 212,402 5.89%

10/29/06 12,024 -212,402 5.66%

6/18/06 15,422 212,402 7.26%

2/23/06 16,615 212,402 7.82%

1/11/06 16,745 212,402* 7.88%

9/27/05 9,418 221,262 4.26%

8/16/05 10,008 221,262 4.52%

8/10/05 12,735 221,262 5.76%

5/11/05 12,482 221,262 5.64%

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page El-20 Acceptance Criteria Percentage of Date Result (sccm) (sccm) Acceptance Criteria 5/4/05 18,183 221,262 8.22%

5/3/05 17,595 221,262 7.95%

3/31/05 17,477 221,262 7.90%

1/27/05 13,882 221,262 6.27%

11/23/04 13,984 221,262 6.32%

9/13/04 11,747 221,262 5.31%

6/29/04 12,262 221,262 5.54%

1/27/04 16,013 221,262 7.24%

1/15/04 15,533 221,262 7.02%

12/11/03 15,533 221,262 7.02%

12/4/03 20,405 221,262 9.22%

• Alternate Source Term adopted.

4.9 NRC Information Notice 92-20, Inadequate Local Leak Rate Testinq NRC Information Notice 92-20 was issued to alert licensees to problems with local leak rate testing of two-ply stainless steel bellows used on piping penetrations at some plants. Specifically, local leak rate testing could not be relied upon to accurately measure the leakage rate that would occur under accident conditions since, during testing, the two plies in the bellows were in contact with each other, restricting the flow of the test medium to the crack locations. Any two-ply bellows of similar construction may be susceptible to this problem.

All ONS 1 piping and ventilation penetrations are of the rigid welded type and are solidly anchored to the Reactor Building wall or foundation slab, thus precluding any requirements for expansion bellows.

4.10 Supplemental Inspection Requirements Prior to initiating a Type A test, a general visual examination of accessible interior and exterior surfaces of the containment system for structural problems that may affect either the containment structure leakage integrity or the performance of the Type A test is performed. This inspection is typically conducted in accordance with the ONS 1 Containment In-service Inspection Plan, which implements the requirements of ASME, Section Xl, Subsection IWE / IWL.

Identification and evaluation of inaccessible areas are addressed in accordance with the requirements of 10 CFR 50.55a(b)(2)(ix)(A) and 10 CFR 50.55a(b)(2)(viii)(E).

Examination of pressure-retaining bolted connections and evaluation of containment bolting flaws or degradation are performed in accordance with the requirements of 10 CFR 50.55a(b)(ix)(G) and 10 CFR 50.55a(b)(ix)(H), as modified by relief request #03-GO-010. Each ten-year ISI interval is divided into three approximately equal-duration inspection periods for IWE, and 24-month periods for IWL examinations and tests (every 5 years).

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-21 Since a 11.25-year ILRT interval spans at least three IWE ISI inspection periods, the frequency of the examinations performed in accordance with the IWE program satisfies the requirement of NEI 94-01, Revision 0, Section 9.2.3.2, to perform the general visual examinations during at least two other outages before the next Type A test, ifthe Type A test interval is to be extended to 11.25 years. Duke Energy intends to credit ASME Section Xl Table IWE-2500-1 Item E1.11 visual exams towards satisfying the requirements of 10 CFR 50, Appendix J. This is in accordance with TS 5.5.2, which requires that "Accessible interior and exterior surfaces of metallic pressure retaining components of the containment system shall be visually examined at least three times every ten years, including during each shutdown for SR 3.6.1.1 Type A test, prior to initiating the Type A test."

Since an 11.25 year ILRT interval spans at least two IWL ISI inspection periods, the frequency of the examinations performed in accordance with the IWL program satisfies the frequency requirement of NEI 94-01, Revision 0, Section 9.2.3.2. However, because the Type A Test may not coincide with scheduled IWL examinations, TS 5.5.2 requires that "Accessible concrete surfaces and post-tensioning system component surfaces of the concrete containment shall be visually examined prior to initiating SR 3.6.1.1 Type A test. These visual examinations, or any portion thereof, shall be performed no earlier than 90 days prior to the start of refueling outages in which Type A tests will be performed." When the Type A Test does not coincide with scheduled IWL examinations, the examination of containment accessible concrete surfaces is performed in accordance with applicable site technical procedures, but is not credited towards satisfying the IWL requirements.

The ASME Code,Section XI, IWE and IWL examination requirements, in conjunction with TS 5.5.2, provide assurance that visual examinations of accessible surfaces of the containment shall be conducted at appropriate frequencies between each Type A test.

4.11 Plant-Specific Confirmatory Analysis The purpose of this analysis is to provide risk insights about extending the currently allowed containment Type A Integrated Leak Rate Test (ILRT) interval by 15 months.

The extended test interval is a one-time 15-month increase over the currently approved 10-year test interval. This translates to an extended test interval of 11.25 years. The extension would allow for substantial cost savings as the ILRT could be deferred for an additional scheduled refueling outage for the Oconee Nuclear Station (ONS). The risk assessment follows the guidelines from NEI 94-01, Revision 2A, the methodology used in EPRI TR-104285 the NEI "Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals" from November 2001, the methodology used for Calvert Cliffs to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval, the methodology used in EPRI 1009325, Revision 2, and the methodology improvements in EPRI 1018243.

The findings of the ONS risk assessment confirm the general findings of previous studies that the risk impact associated with extending the ILRT interval from ten years to 11.25 years is "small." The ONS plant-specific results for extending ILRT interval from the current 10 years to 11.25 years are summarized below:

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-22

" Since the ILRT does not impact Core Damage Frequency (CDF), the relevant criterion is Large Early Release Frequency (LERF). The increase in LERF resulting from a change in the Type A ILRT test interval from three in 10 years to one in 11.25 years is very conservatively estimated to be "small."

• An additional assessment of the impact from external events was also performed. In this sensitivity case, the change in the total ONS LERF (including external events) was conservatively estimated to be "small." Similar sensitivity analysis of internal flood events were also performed and resulted in the same conclusions. As such, the estimated change in LERF from sensitivity studies is also determined to be "small."

• The change in Type A test frequency to one per 11.25 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.021 person-rem/year. EPRI Report No. 1009325, Revision 2-A states that a very small population dose is defined as an increase of < 1.0 person-rem per year, or < 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. Moreover, the risk impact when compared to other severe accident risks is "negligible."

• The increase in the conditional containment failure from the three in 10 year interval to one in 11.25 year interval is 0.634%. EPRI Report No. 1009325, Revision 2-A, states that increases in CCFP of < 1.5 percentage points is very small. Therefore, this increase is judged to be "very small."

Therefore, increasing the ILRT interval to 11.25 years is considered to be insignificant since it represents a "very small" change to the ONS risk profile.

The NRC, in NUREG-1493, has previously concluded that:

• Reducing the frequency of Type A tests (ILRTs) from three per 10 years to one per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B or Type C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements.

" Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integrated leakage-rate tests is possible with minimal impact on public risk. The impact of relaxing the ILRT frequency beyond one in 20 years has not been evaluated. Beyond testing the performance of containment penetrations, ILRTs also test integrity of the containment structure.

The findings for ONS confirm these general findings on a plant-specific basis considering the severe accidents evaluated for ONS, the ONS containment failure modes, and the local population surrounding ONS.

Enclosure 1 - Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-23 The insights from this risk analysis support the deterministic analysis showing that there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner of this license request.

5. REGULATORY EVALUATION 5.1 Siqnificant Hazards Consideration A change is proposed to the Oconee Nuclear Station, Unit 1 (ONS 1), Technical Specification 5.5.2, "Containment Leakage Rate Testing Program." The proposed amendment would extend the Type A test required by TS 5.5.2 for Unit 1 by approximately 15 months.

Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

The proposed exemption involves a one-time extension to the current interval for ONS 1 Type A containment testing. The current test interval of 120 months (10 years) would be extended on a one-time basis to no longer than approximately 135 months from the last Type A test. The proposed extension does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.

This proposed extension is for next ONS 1 Type A containment leak rate test only. The Type B and C containment leak rate tests would continue to be performed at the frequency currently required by the ONS TS. As documented in NUREG 1493, Type B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The ONS 1 Type A test history supports this conclusion.

The integrity of the containment is subject to two types of failure mechanisms that can be categorized as (1) activity based and (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-24 that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with ASME Section Xl, the Maintenance Rule, and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test.

Based on the above, the proposed extension does not involve a significant increase in the consequences of an accident previously evaluated. Therefore, it is concluded that the proposed amendment does not significantly increase the 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 amendment to the TS involves a one-time extension to the current interval for the ONS 1 Type A containment test. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled.

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

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

Response: No.

The proposed amendment to the TS involves a one-time extension to the current interval for the ONS 1 Type A containment test. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained.

The proposed change involves only the extension of the interval between Type A containment leak rate tests for ONS 1. The proposed surveillance interval extension is bounded by the 15-month extension currently authorized within NEI 94-01, Revision

0. Type B and C containment leak rate tests would continue to be performed at the frequency currently required by TS. Industry experience supports the conclusion that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section XI, TS

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-25 and the Maintenance Rule serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing.

The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A test interval.

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

Based on the above, Duke Energy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Requlatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met.

10 CFR 50.54(o) requires primary reactor containments for water-cooled power reactors to be subject to the requirements of Appendix J to 10 CFR 50, "Leakage Rate Testing of Containment of Water Cooled Nuclear Power Plants." Appendix J specifies containment leakage testing requirements, including the types required to ensure the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. In addition, Appendix J discusses leakage rate acceptance criteria, test methodology, frequency of testing and reporting requirements for each type of test.

RG 1.163 was developed to endorse NEI 94-01, Revision 0 with certain modifications and additions.

The adoption of the Option B performance-based containment leakage rate testing for Type A testing did not alter the basic method by which Appendix J leakage rate testing is performed; however, it did alter the frequency at which Type A, Type B, and Type C containment leakage tests must be performed. Under the performance-based option of 10 CFR 50, Appendix J, the test frequency is based upon an evaluation that review "as-found" leakage history to determine the frequency for leakage testing which provides assurance that leakage limits will be maintained. The change to the Type A test frequency did not directly result in an increase in containment leakage. Similarly, the proposed change to the Type A test frequency will not directly result in an increase in containment leakage.

Based on the considerations above, (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 continue to be conducted in accordance with the site licensing basis, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-26 5.3 Precedent This request is similar in nature to the following license amendments authorized by the NRC:

• December 29, 1994 (ML011080782), for Nine Mile Point Nuclear Station Unit 1,

• June 2, 2003 (ML031320686), for Vermont Yankee Nuclear Power Station,

• July 20, 2009 (ML091540158) for Arkansas Nuclear One, Unit No. 2, and

• August 23, 2010 (ML102090137) for Palisades Nuclear Plant.

5.4 Conclusion In conclusion, Duke Energy has determined that the proposed change does not require any exemptions or relief from regulatory requirements, other than the TS, and does not affect conformance with any regulatory requirements / criteria.

6. ENVIRONMENTAL CONSIDERATION The proposed amendment 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 amendment 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 amendment.

7. REFERENCES

1. ONS Updated Final Safety Analysis Report - 31 Dec 2010

2. ONS Technical Specifications

3. 10 CFR 50 Appendix J, "Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors"

4. Nuclear Energy Institute, NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J"

5. Nuclear Energy Institute, NEI 94-01, Revision 2-A," Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J,"

October 2008

6. NUREG-1493, "Performance-Based Containment Leak-Test Program"

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-27

7. Electric Power Research Institute, EPRI TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals"

8. Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program" September 1995

9. ASME Boiler and Pressure Vessel Code, Section Xl, 1992 Edition with the 1992 Addenda.

10. Document #O-ISIC2-62-0001, "Second Interval Containment Inservice Inspection Plan" Revision 6

11. Procedure QA-516, "Evaluation of ISI Indications"

12. Duke Power Company Mechanical Systems Engineering Support Program For 10 CFR Part 50-Appendix J, Revision 1

13. Nuclear System Directive: 318, Coating Program, Revision 5

14. Regulatory Guide 1.54, Service Level 1,11, and III Protective Coatings Applied to Nuclear Power Plants, Revision 2

15. 10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

16. American National Standards Institute, ANSI N101.2, "Protective Coatings (Paints) for Light Water Nuclear Reactor Containment Facilities," 1972

17. Regulatory Guide 1.82, Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident, Revision 3

18. PT/1/A/0150/034, "Leak Rate History And Reactor Building Leak Rate Verification Results"

19. NRC Information Notice 92-20, Inadequate Local Leak Rate Testing

20. Electric Power Research Institute, EPRI Report No. 1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Interval, October 2008

21. ML011080782, Issuance Of Amendment For Nine Mile Point Nuclear Station Unit No. 1 (TAC NO. M90278) December 29, 1994

22. ML031320686, Vermont Yankee Nuclear Power Station - Issuance Of Amendment Re: One-Time Extension Of Appendix J Type A Integrated Leakage Rate Test Interval (TAC NO. MB6507) June 2, 2003

- Evaluation of Proposed Changes License Amendment Request No. 2012-03 April 3, 2012 Page E1-28

23. ML091540158, Arkansas Nuclear One, Unit No.2 -Issuance Of Amendment Re:

One-Time Extension To 10-Year Frequency Of Integrated Leak Rate Test (TAC NO. MD9502) July 20, 2009

24. ML102090137, Palisades Nuclear Plant -Issuance Of Amendment Re: One-Time Extension To The Integrated Leak Rate Test Interval (TAC NO. ME2122)

August 23, 2010

25. ML012050049, Issuance Of Technical Specification Amendments -Oconee Nuclear Station, Units 1, 2, And 3 (TAC NOS. M96317, M96318, M96319)

26. ML11186A906, Oconee Nuclear Station, Units 1, 2, And 3, Issuance Of Amendments Regarding A Proposed Change To The Technical Specifications To Adopt Technical Specification Task Force (TSTF) Technical Change Traveler 52, Revision 3, To Implement Option B Of Appendix J To Title 10 Of The Code Of Federal Regulations, Part 50 (TAC Nos. ME455, ME4558, And ME4559)

Attachment 1 Proposed Technical Specification Changes (mark-up)

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained.

5.5.1 Offsite Dose Calculation Manual (ODCM)

The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program.

Licensee initiated changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained.

This documentation shall contain:

1. sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and

2. a determination that the change(s) do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;

b. Shall become effective after the approval of the Station Manager; and

c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

5.5.2 Containment Leakage Rate Testing Program INSERT: The next A program shall establish the leakage rate testing of the containment as required Unit 1 ILRT by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by following the approved exemption This program shall be in accordance with the guidelines December 8, 2003 contained in atory Guide 1.163, "Performance-Based Containment Leak-test shall be performed no later than March 8, 2015.

OCONEE UNITS 1, 2, & 3 5.0-7 Amendment Ns.t 37-7, &

Programs and Manuals 5.5 5.5 Programs andManuals 5.5.2 Containment Leakage Rate Testing Program (continued)

Test Program," dated September 1995. Containment system visual examinations required by Regulatory Guide 1.163, Regulatory Position C.3 shall be performed as follows:

1. Accessible concrete surfaces and post-tensioning system component surfaces of the concrete containment shall be visually examined prior to initiating SR 3.6.1.1 Type A test. These visual examinations, or any portion thereof, shall be performed no earlier than 90 days prior to the start of refueling outages in which Type A tests will be performed. The validity of these visual examinations will be evaluated should any event or condition capable of affecting the integrity of the containment system occur between the completion of the visual examinations and the Type A test.

2. Accessible interior and exterior surfaces of metallic pressure retaining components of the containment system shall be visually examined at least three times every ten years, including during each shutdown for SR 3.6.1.1 Type A test, prior to initiating the Type A test.

The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 59 psig. The containment design pressure is 59 psig.

The maximum allowable containment leakage rate, La, at Pa, shall be 0.20% of the containment air weight per day.

Leakage rate acceptance criterion is:

a. Containment leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests, and < 0.75 La for Type A tests; The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

Nothing in these Technical Specifications shall be construed to modify the testing Frequencies of 10 CFR 50, Appendix J.

OCONEE UNITS 1, 2, & 3 5.0-8 Amendment N&

Attachment 2 Proposed Technical Specification Changes (retype)

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained.

5.5.1 Offsite Dose Calculation Manual (ODCM)

The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program.

Licensee initiated changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained.

This documentation shall contain:

1. sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and

2. a determination that the change(s) do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;

b. Shall become effective after the approval of the Station Manager; and

c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

5.5.2 Containment Leakage Rate Testing Program A program shall establish 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. The next Unit 1 ILRT following the December 8, 2003 test shall be performed no later than March 8, 2015. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, OCONEE UNITS 1, 2, & 3 5.0-7 Amendment Nos. XXX, XXX, & XXX

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.2 Containment Leakage Rate Testing Program (continued)

"Performance-Based Containment Leak-Test Program," dated September 1995. Containment system visual examinations required by Regulatory Guide 1.163, Regulatory Position C.3 shall be performed as follows:

3. Accessible concrete surfaces and post-tensioning system component surfaces of the concrete containment shall be visually examined prior to initiating SR 3.6.1.1 Type A test. These visual examinations, or any portion thereof, shall be performed no earlier than 90 days prior to the start of refueling outages in which Type A tests will be performed. The validity of these visual examinations will be evaluated should any event or condition capable of affecting the integrity of the containment system occur between the completion of the visual examinations and the Type A test.

4. Accessible interior and exterior surfaces of metallic pressure retaining components of the containment system shall be visually examined at least three times every ten years, including during each shutdown for SR 3.6.1.1 Type A test, prior to initiating the Type A test.

The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 59 psig. The containment design pressure is 59 psig.

The maximum allowable containment leakage rate, La, at Pa, shall be 0.20% of the containment air weight per day.

Leakage rate acceptance criterion is:

a. Containment leakage rate acceptance criterion is _<1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are _< 0.60 La for the Type B and C tests, and < 0.75 La for Type A tests; The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

Nothing in these Technical Specifications shall be construed to modify the testing Frequencies of 10 CFR 50, Appendix J.

OCONEE UNITS 1, 2, & 3 5.0-8 Amendment Nos. XXX, XXX, & XXX I