AEP-NRC-2014-09, License Amendment Request to Revise Technical Specification Section 5.5.14, Containment Leakage Rate Testing Program

License Amendment Request to Revise Technical Specification Section 5.5.14, Containment Leakage Rate Testing Program

ML14071A435
Person / Time
Site:	Cook
Issue date:	03/07/2014
From:	Gebbie J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2014-09
Download: ML14071A435 (29)

Text

INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant One Cook Place POWER Bridgman, MI 49106 A unit of American Electric Power IndianaMichiganPower.com March 7, 2014 AEP-NRC-2014-09 10 CFR 50.90 Docket Nos.: 50-315 50-316 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555-0001 Donald C. Cook Nuclear Plant Units 1 and 2 LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATION SECTION 5.5.14, "CONTAINMENT LEAKAGE RATE TESTING PROGRAM."

References:

1. Nuclear Energy Institute 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 2012, ADAMS Accession No. ML .12221A202.

2. Electrical Power Research Institute (EPRI) report, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325" (EPRI product No. 1018243).

3. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, May 2011, ADAMS Accession No. ML13109A112.

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company .(I&M), licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, is submitting a request for an amendment to the Technical Specifications (TS) for CNP Unit 1 and Unit 2.

This application for amendment to the CNP TS proposes to revise Section 5.5.14, "Containment Leakage Rate Testing Program", by adopting Nuclear Energy Institute (NEI) 94-01 Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J" (Reference 1) as the implementation document for the performance-based Option B of 10 CFR Part 50, Appendix J. This revision is for:

1) A permanent 15-year interval for Type A containment integrated leak rate testing (ILRT);

2) An extended 75-month interval for Type C local leakage rate tests.

The proposed amendment is considered risk-informed. An evaluation has been performed to assess the risk impact of the proposed change. The risk assessment follows the guidelines of

U.S. Nuclear Regulatory Commission AEP-NRC-2014-09 Page 2 Reference 1, and the corresponding Electrical Power Research Institute (EPRI) report, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325" (EPRI Product No. 1018243) (Reference 2), and the guidance of Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." (Reference 3) to this letter provides an affirmation statement. Enclosure 2 is a safety evaluation of the proposed change to Section 5.5.14 of the Administrative Section of the TSs. Enclosure 3 is a risk impact assessment for permanently extending containment Type A test interval. Enclosures 4 and 5 are marked up copies of the applicable Unit 1 and Unit 2 TS pages, respectively, provided for information purposes. Enclosure 6 contains documentation of the technical adequacy of the Probabilistic Risk Assessment.

I&M requests review and approval of this application by March 1, 2015, in order to incorporate these changes into the CNP schedule. The license amendment will be implemented within 60 days of U.S. Nuclear Regulatory Commission approval. This would allow deferral of the next ILRT Type A tests, currently scheduled for the Unit 1 Cycle 27 refueling outage in the spring of 2016 and the Unit 2 Cycle 23 refueling outage in the fall of 2016. The last ILRT was completed on April 22, 2006, (Unit 2) and November 1, 2006, (Unit 1). CNP intends to use six months of the 15 months allowed for Type A interval extensions per NEI 94-01 Revision 0, Section 9.1 if this amendment is not approved. This would allow appropriate time for test planning and preparations if required.

The revision is to allow ILRT within 15 years from the last test on each unit. This application represents a cost-beneficial licensing change. The ILRT imposes a significant expense on the station while the safety benefit of performing it within 10 years versus 15 years is minimal.

Copies of this letter and its enclosures are being transmitted to the Michigan Public Service Commission and Michigan Department of Environmental Quality, in accordance with the requirements of 10 CFR 50.91.

There are no new regulatory commitments made in this letter. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.

Sincerely, Ho .Pýe bb~ite Site Vice President JMT/amp

U.S. Nuclear Regulatory Commission AEP-NRC-2014-09 Page 3

Enclosures:

1. Affirmation

2. Safety Evaluation of Proposed Changes to Section 5.5.14

3. Donald C. Cook Nuclear Plant Calculation, PRA-ILRT "Risk Impact Assessment For Permanently Extending Containment Type A Test Interval"

4. Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages Marked to Show Proposed Changes

5. Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages Marked to Show Proposed Changes

6. Probabilistic Risk Assessment Technical Adequacy c: J. T. King- MPSC MDEQ - RMD/RPS NRC Resident Inspector C. D. Pederson, NRC Region III T. J. Wengert- NRC Washington DC A. J. Williamson, AEP Ft. Wayne, w/o attachments

Enclosure 1 to AEP-NRC-2014-09 AFFIRMATION I, Joel P. Gebbie, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company Joel P. Gebbie Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS DAY OF 2014 DANIELLE BURGOYNE Notary c* \c Notary Public. State of Michigan County of Berrien My Commission Expires 04-04-2018 Acting In the County of c- -

My Commission Expires \ ,- - -ZZ\,

Enclosure 2 to AEP-NRC-2014-09 Safety Evaluation of Proposed Changes to Section 5.5.14

4.0 TECHNICAL ANALYSIS

4.1 Description of Containment 4.2 Integrated Leak Rate Test History 4.3 Type B and C Testing Programs 4.4 Supplemental Inspection Requirements 4.4.1 IWE Examinations 4.4.2 IWL Examinations 4.5 Deficiencies Identified 4.6 Plant-Specific Confirmatory Analysis 4.6.1 Methodology 4.6.2 PRA Technical Adequacy 4.6.3 Conclusion of Plant-Specific Risk Assessment Results 5.0 REGULATORY ASSESSMENT 5.1 Applicable Regulatory Requirements/Criteria 5.2 No Significant Hazards Consideration 5.3 Environmental Considerations

6.0 CONCLUSION

7.0 PRECEDENCE

8.0 REFERENCES

to AEP-NRC-2014-09 Page 2 This license amendment request (LAR) is to amend Operating License Numbers DPR-58 and DPR-74 for Donald C. Cook Nuclear Plant (CNP) Unit 1 and Unit 2, respectively. Documents referenced in this enclosure are identified in Section 8.0.

1.0 DESCRIPTION

Indiana Michigan Power Company (I&M) requests an amendment to the CNP Unit 1 Operating License DPR-58 and Unit 2 Operating License DPR-74 by incorporating the proposed change for the CNP Units 1 and 2 Technical Specifications (TS). The proposed change is a request to revise TS 5.5.14, "Containment Leakage Rate Testing Program" for CNP Units 1 and 2.

I&M requests review and approval of this application by March 1, 2015, in order to incorporate these changes into our plant schedule. The license amendment will be implemented within 60 days of the issuance of the license amendment. This would allow extending the interval prior to the next integrated leak rate testing (ILRT) Type A tests, currently scheduled for the Unit 1 Cycle 27 refueling outage in the spring of 2016 and the Unit 2 Cycle 23 refueling outage in the fall of 2016. The last ILRT was completed on April 22, 2006, (Unit 2) and November 1, 2006, (Unit 1). CNP intends to use 6 months of the 15 months allowed for Type A interval extensions per Nuclear Energy Institute (NEI) 94-01 Revision 0, Section 9.1 if this amendment is not approved. This would allow appropriate time for test planning and preparations if required.

The proposed amendment would revise the CNP Unit 1 and Unit 2 TS 5.5.14 by replacing the reference to Regulatory Guide (RG) 1.163 with a reference to NEI 94-01, Revision 3-A, as the implementation document used by CNP Unit 1 and Unit 2 performance-based leakage testing program in accordance with Option B of Title 10 of the Code of Federal Regulations (CFR)

Part 50. The proposed amendment would allow the next primary containment ILRT to be performed within 15 years from the last ILRT as opposed to the current 10-year interval, and would allow successive ILRTs to be performed at 15-year intervals.

Additionally, the proposed amendment would revise the maximum interval for the performance of Type C local leakage rate tests (LLRT) to a 75-month interval, under 10 CFR Part 50, Appendix J.

2.0 PROPOSED CHANGE

TS Section 5.5.14, Containment Leakage Rate Testing Program, currently states the following:

"a. 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. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:

1. The Type A testing Frequency specified in NEI 94-01, Revision 0, Paragraph 9.2.3, as "at least once per 10 years based on acceptable performance history" is modified to be "at least once per 15 years based on acceptable performance history." This change applies only to the interval following the Type A test performed in October 1992.

to AEP-NRC-2014-09 Page 3

2. A one-time exception to the requirement to perform post-modification Type A testing is allowed for the steam generators and associated piping, as components of the containment barrier. For this case, American Society of Mechanical Engineers (ASME) Section Xl leak testing will be used to verify the leak tightness of the repaired or modified portions of the containment barrier.

Entry into MODES 3 and 4 following the extended outage that commenced in 1997 may be made to perform this testing."

This proposed change in the current licensing basis is a permanent change of the Type A test interval from 10 years to 15 years and of the Type C tests for up to 75 months. The approved change of the ILRT and LLRT would be incorporated into TS 5.5.14 by: modifying the existing paragraph "a" to reflect the new revision of NEI 94-01, and deleting subparagraphs "1" and "2",

as follows:

a. "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 NRC endorsements. This program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A, "Industry Guidelines for Implementing Performance-Based Option of 10 CFR 50, Appendix J," dated July 2012." contains the existing Unit 1 TS Page 5.5-14 marked up to show the proposed changes to TS 5.5.14. Enclosure 5 contains the existing Unit 2 TS Page 5.5-14 marked up to show the proposed changes to TS 5.5.14. New clean Unit 1 and Unit 2 TS pages, with proposed changes incorporated, will be provided to the U. S. Nuclear Regulatory Commission (NRC) Licensing Project Manager when requested.

3.0 BACKGROUND

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 TS 5.5.14. The testing requirements assure that periodic surveillance of containment penetrations and isolation valves is performed so that proper maintenance and repairs are performed on the systems and components penetrating containment during the service life of the 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 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 containment penetrations; and (3) Type C tests, intended to measure containment isolation (CI) valve leakage. 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. This request modifies the existing Appendix J Type A, Type B, and Type C testing intervals but does not change the Appendix J Type A, Type B, or Type C test methods. The ILRT imposes significant expense on the station while the safety benefit of performing it within 10 years, versus 15 years, is minimal. This request represents a cost-beneficial licensing change with no reduction in safety margin.

to AEP-NRC-2014-09 Page 4 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 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. Also in 1995, RG 1.163 was issued. The RG endorsed NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 94-01, 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 from the containment Type A (ILRT) test from three tests in ten years to one test in ten years. This relaxation was based on an NRC risk program and Electric Power Research Institute (EPRI) Topical Report (TR)-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals," both of which illustrated that the risk increase associated with extending the ILRT surveillance interval was very small.

NEI 94-01, Revision 2, describes an approach for implementing the optional performance-based requirements of Option B described in 10 CFR 50, Appendix J, which includes provisions for extending Type A intervals to up to 15 years and incorporates the regulatory positions stated in RG 1.163. It delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. This method uses industry performance data, plant-specific performance data, and risk insights in determining the appropriate testing frequency. NEI 94-01, Revision 2, also discusses the performance factors that licensees must consider in determining test intervals. However, it does not address how to perform the tests because these details are included in existing documents (e.g., American National Standards Institute/American Nuclear Society [ANSI/ANS]-56.8-2002). The NRC final Safety Evaluation (SE), issued by letter dated June 25, 2008, documents the NRC's evaluation and acceptance of NEI 94-01, Revision 2, subject to the specific limitations and conditions listed in Section 4.1 of the SE. The accepted version of NEI 94-01 has subsequently been issued as Revision 2-A dated October 2008.

EPRI TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals," Revision 2, provides a risk impact assessment for optimized ILRT intervals of up to 15 years, utilizing current industry performance data and risk-informed guidance, primarily Revision 1 of RG 1.174, "An approach for using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Bases." The NRC's final SE, issued by letter dated June 25, 2008, documents the NRC's evaluation and acceptance of EPRI TR-104285, Revision 2, subject to the specific limitations and conditions listed in Section 4.2 of the SE. An accepted version of EPRI TR-1009325 was subsequently issued as Revision 2-A (also identified as TR-1 018243) dated October 2008.

NEI 94-01, Revision 3, describes an approach for implementing the optional performance-based requirements of Option B described in 10 CFR 50, Appendix J, which includes provisions for extending Type A and Type C intervals to up to 15 years and 75 months, respectively, and incorporates the regulatory positions stated in RG 1.163. It delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance to AEP-NRC-2014-09 Page 5 testing frequencies. This method uses industry performance data, plant-specific performance data, and risk insights in determining the appropriate testing frequency. NEI 94-01, Revision 3, also discusses the performance factors that licensees must consider in determining test intervals. However, it does not address how to perform the tests because these details are included in existing documents (e.g., ANSI/ANS-56.8-2002). The NRC final SE issued by letter dated June 8, 2012, documents the NRC's evaluation and acceptance of NEI 94-01, Revision 3, subject to the specific limitations and conditions listed in Section 4.0 of the SE. The accepted version of NEI 94-01 has subsequently been issued as Revision 3-A dated July 2012.

EPRI TR-109325, Revision 2, provides a validation of the risk impact assessment of EPRI TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals," dated August 1994. The assessment validates increasing allowable extended LLRT intervals to the 120 months as specified in NEI 94-01, Revision 0. However, the industry requested that the allowable extended interval for Type C LLRTs to be increased only to 75 months, to be conservative, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total). The NRC final SE issued by letter dated June 8, 2012, documents the NRC's evaluation and acceptance of EPRI TR-1009325 as a validation of EPRI-104285, Revision 2 bases to extend Type C LLRT to 75 months, subject to the specific limitations and conditions listed in Section 4.0 of the SE.

4.0 TECHNICAL ANALYSIS

As required by 10 CFR 50.54(o), the CNP Unit 1 and Unit 2 containments are subject to the requirements set forth in 10 CFR 50, Appendix J. Option B of Appendix J requires that test intervals for Type A, Type B, and Type C testing be determined by using a performance-based approach. Currently, the CNP 10 CFR 50 Appendix J Testing Plan is based on RG 1.163, which endorses NEI 94-01, Revision 0. This LAR proposes to revise the CNP 10 CFR 50, Appendix J Testing Plan by implementing the guidance in NEI 94-01, Revision 3-A.

In the SE issued by the NRC dated June 8, 2012, the NRC concluded that NEI 94-01, Revision 3-A, as modified to include two limitations and conditions, is acceptable for referencing by licensees proposing to amend their TS with regard to containment leakage rate testing for the optional performance-based requirements of Option B of 10 CFR 50, Appendix J.

The following addresses each of the limitations and conditions of the 2008 and 2012 SEs.

Limitation / Condition CNP Response (from Section 4.1 of SE dated June 25, 2008

1. For calculating the Type A leakage rate, the licensee Following the NRC approval of this LAR, CNP will use should use the definition in the NEI 94-01, Revision the definition in Section 5.0 of NEI 94-01, Revision 3-A, 2, in lieu of that in ANSI/ANS-56.8-2002. for calculating the Type A leakage rate when future CNP Type A tests are performed. The definition in Revision 2-A and 3-A are identical.

2. The licensee submits a schedule of containment A schedule of containment inspections is provided in inspections to be performed prior to and between Section 4.4 below.

Type A tests.

3. The licensee addresses the areas of the General visual examination of accessible interior and containment structure potentially subjected to exterior surfaces of the containment system for structural degradation. problems is conducted in accordance with the CNP IWE/IWL Containment Inservice Inspection Plans which to AEP-NRC-2014-09 Page 6 implement the requirements of the ASME,Section XI, Subsections IWE and IWL, as required by 10 CFR 50.55a(a).

4. The licensee addresses any test and inspections CNP has already replaced the steam generators and the performed following major modifications to the reactor vessel closure heads. The design change containment structure, as applicable, process addressed the testing requirements. Similarly, any future containment modifications will be addressed by the site design change process.

5. The normal Type A test interval should be less than I&M acknowledges and accepts this NRC staff position, 15 years. If a licensee has to utilize the provisions as communicated to the nuclear industry in Regulatory of Section 9.1 of NEI 94-01, Revision 2, related to Issue Summary 2008-27 dated December 8, 2008.

extending the ILRT interval beyond 15 years, the licensee must demonstrate to the NRC staff that it is an unforeseen emergent condition.

6. For plants licensed under 10 CFR Part 52, Not applicable. CNP Units 1 and 2 are not licensed applications requesting a permanent extension of pursuant to 10 CFR Part 52.

the ILRT surveillance interval to 15 years should be deferred until after the construction and testing of containments for that design have been completed and applicants have confirmed the applicability of NEI 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, including the use of past containment ILRT data.

Limitation I Condition CNP Response (from Section 4.1 of SE dated July 2012)

The staff is allowing the extended interval for Type C Following approval of the amendment, CNP will follow LLRTs to be increased to 75 months with the the guidance of NEI 94-01, Revision 3-A to assess and requirement that a licensee's post-outage report monitor margin between the Type B and C leakage rate include the margin between the Type B and Type C summation and the regulatory limit. This will include leakage rate summation and its regulatory limit. In corrective actions to restore margin to an acceptable addition, a corrective action plan shall be developed level.

to restore the margin to an acceptable level. The staff is also allowing the non-routine emergent extension out to 84 months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI 94-01, Revision 3-A. At no time shall an extension be allowed for Type C valves that are restricted categorically (e.g., BWR Main Steam Isolation Valve), and those valves with a history of leakage, or any valves held to either a less than maximum interval or to the base refueling cycle interval. Only non-routine emergent conditions allow an extension to 84 months.

to AEP-NRC-2014-09 Page 7

2. When routinely scheduling any LLRT valve interval Following the approval of the amendment, consistent beyond 60 months and up to 75 months, the primary with the guidance of Section 11.3.2 of NEI 94-01, containment leakage rate testing program trending Revision 3-A CNP will estimate the amount of or monitoring must include an estimate of the understatement in the Types B and C total and include amount of understatement in the Types B and C determination of the acceptability in a post-outage report.

total and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

4.1 Description of Containment The steel-lined, reinforced concrete containment structure, including foundations, access hatches, and penetrations is designed and constructed to maintain full containment integrity when subject to accident temperatures and pressure, and the postulated earthquake conditions.

The structure consists of side walls measuring 113 feet (ft.) (nominal) in height from the liner on the base to the spring line of the dome and has a nominal inside diameter of 115 ft.. The cylinder is 4 ft. - 6 inch (in.) thick at the base, tapering to 3 ft. - 6 in. seven feet above the base.

The thickness of the cylinder remains at 3 ft. - 6 in. to the dome spring line. The thickness of the dome is 3 ft. - 6 in. at the spring line tapering uniformly to 2 ft. - 6 in. at the peak of the dome. The base mat consists of a 10 ft. thick structural concrete slab, increasing to 20 ft.

adjacent to the recirculation sump area.

The basic structural elements considered in the design of the containment structure are the base slab, the vertical cylinder and the hemispherical dome, all acting as one structure. The vertical cylindrical wall and the dome of the steel liner are anchored to the concrete by means of horizontal and vertical stiffener angles. In addition, Nelson studs welded to the stiffener angles extend into the concrete and are anchored behind the first layer of reinforcing, thereby preventing pull-out in case of local concrete cracking. The steel base liner is anchored to the concrete by welding it to continuous steel tee bars which in turn are welded to structural members anchored into the base mat. The base liner is covered by a 2 ft. - 0 in. concrete mat.

The underground portion of the containment vessel is waterproofed in order to prevent possible corrosion of the reinforcing steel and liner plate due to seepage of ground water. The waterproofing consists of a continuous impervious membrane, which is placed under the mat, and on the outside of the walls. The membrane placed under the mat extends up and around the walls and is taped to the membrane placed on the outside of the walls, thus providing a continuous waterproof surface.

The reinforced concrete structure was designed in accordance with the applicable portions of the American Concrete Institute (ACI) codes ACI-318-63 and ACI-301-66. The structural steel components were designed in accordance with the American Institute of Steel Construction, AISC-69 specifications.

to AEP-NRC-2014-09 Page 8 The containment is divided into three main compartments. These are:

a) The lower compartment.

b) The upper compartment.

c) The ice condenser compartment.

The lower compartment encloses the reactor system and associated auxiliary systems equipment. The upper compartment contains the refueling cavity, refueling equipment and polar crane used during refueling and maintenance operations. The upper and lower compartments are separated by a divider barrier. The ice condenser, which contains borated ice provided to absorb the loss of-coolant accident (LOCA) energy, is in the form of an enclosed and refrigerated annular compartment, located circumferentially between the crane wall and the outer wall of the containment and extends from below to above the operating deck and divider barrier.

The reactor containment structure is a reinforced concrete vertical right cylinder with a slab base and a hemispherical dome. A welded steel liner with a nominal thickness of 3/8 in. at the dome and wall, and 1/4 in. at the bottom is attached to the inside face of the concrete shell, to insure a high degree of leak tightness. The containment structure is designed to contain the radioactive material, which might be released, following a LOCA. The structure serves as both a biological shield and a pressure container.

The ice condenser is a completely enclosed annular compartment located around, approximately 300 degrees, of the perimeter of the upper compartment of the containment, but penetrating the operating deck so that a portion extends into the containment lower compartment. The lower portion has a series of hinged doors that are exposed to the atmosphere of the lower containment compartment which, for normal plant operation, are designed to remain closed. At the top of the ice condenser is another set of doors that are exposed to the atmosphere of the upper compartment; these also remain closed during normal plant operation. Intermediate deck doors are located below the top deck doors. These doors form the floor of a plenum at the upper part at the Ice condenser and remain closed during normal plant operation. In the ice condenser, ice is held in baskets arranged to promote heat transfer to the ice. A refrigeration system maintains the ice in the solid state. Suitable insulation surrounding both the ice condenser volume and the refrigeration ducts serves to minimize the heat transfer to the ice condenser boundaries.

In the event of a LOCA or steam line break in the containment, the pressure rises in the lower compartment and the door panels located below the operating deck (a portion of the divider barrier) open. This allows the air and steam to flow from the lower compartment into the ice condenser. The resulting pressure increase within the ice condenser causes the intermediate deck doors and the door panels at the top of the ice condenser to open, allowing the air to flow out of the ice condenser into the upper compartment. Steam entering the ice condenser compartment is condensed by the ice, thus limiting the peak pressure and temperature buildup in the containment. Condensation of steam within the ice condenser results in a continual flow of steam from the lower compartment to the condensing surface of the ice, thus reducing the lower compartment pressure. The divider barrier separates the upper and lower compartments and ensures that the steam is directed into the bottom of the ice condenser. Only a limited amount of steam can bypass the ice condenser through the divider barrier.

to AEP-NRC-2014-09 Page 9 The containment liner is enclosed within the containment and thus is not directly exposed to the temperature of the environs. The containment ambient temperature during operation is between 60 and 120 Degrees Fahrenheit (OF), and the ice bed operating temperature is approximately 10-20 0 F, which is above the Nil Ductility Transition Temperature + 30°F for the liner material.

4.2 Integrated Leak Rate Test History Previous ILRT testing confirmed that CNP containment structures' leakage is acceptable, with considerable margin, with respect to the TS acceptance criteria of 0.25 percent (%) of containment air weight at the design basis LOCA pressure (La). Since the last three CNP Units 1 and 2 Type-A results meet the performance leakage rate criteria from NEI 94-01, Revision 3-A, a test frequency of at least once per 15 years would be in accordance with NEI 94-01, Revision 3-A.

Unit 1 Test Date As-Found Leakage Acceptance Limit*

June 1989 Mass point Upper Confidence 0.419 of La 1.0 La Limit (UCL) leakage with penalties:

October 1992 Mass point UCL leakage with 0.044 of La 1.0 La penalties:

November 2006 Mass point UCL leakage with 0.336 of La 1.0 La penalties:

• The total allowable "as-left" leakage is 0.75 La, (La, 0.25% of primary containment air by weight per day, is the leakage assumed in design basis accident radiological analyses) with 0.6 La, the maximum leakage from Type B and C components.

Unit 2 Test Date As-Found Leakage Acceptance Limit*

February 1989 Mass point UCL leakage with 0.075 of La 1.0 La penalties:

May 1992 Mass point UCL leakage with 0.109 of La 1.0 La penalties:

April 2006 Mass point UCL leakage with 0.234 of La 1.0 La penalties:

• The total allowable "as-left" leakage is 0.75 La, (La, U.25% ot primary containment air by weight per day, is the leakage assumed in design basis acicident radiological analyses) with 0.6 La, the maximum leakage from Type B and C components.

Type B and C containment penetrations tests (e.g., electrical penetrations, airlocks, hatches, flanges, and valves) are being performed in accordance with Option B of 10 CFR 50, to AEP-NRC-2014-09 Page 10 Appendix J. The current total penetration leakage on a minimum path basis is less than 20% of the leakage allowed for containment integrity.

No modifications that require a Type A test are planned prior to Ul C30 (fall 2020) and U2C26 (spring 2021), when the next Type A tests will be performed in accordance with 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 either Unit 1 or Unit 2 containments which could have adversely affected containment integrity. There is no anticipated addition or removal of plant hardware within either Unit 1 or Unit 2 containments which could affect leak-tightness.

4.3 Type B and Type C Testing Programs CNP Units 1 and 2 Appendix J, Type B and Type C leakage rate test 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 5.5.14. The Type B and Type C testing program consists of LLRT of penetrations with a resilient seal, double gasketed manways, hatches and flanges, and Cl valves that serve as a barrier to the release of the post-accident containment atmosphere.

A review of the most recent Type B and Type C test results and a comparison with the allowable leakage rate was performed. The combined Type B and Type C leakage has remained below La (110,219 standard cubic centimeters (sccm)) for CNP Units 1 and 2 with the exception of the U2C20 As-Found Minimum Pathway Leakage Rate. In that case, an evaluation was conducted and work was performed to replace the hinge ring retaining dowel pins which were determined to be too short. The maximum and minimum pathway leak rate summary totals for the last three refueling outages are shown below.

Unit 1 April 2010 As-Found Min Pathway Leakage 41,616 sccm As-Left Max Pathway Leakage 28,481 sccm September 2011 As-Found Min Pathway Leakage 46,107 sccm As-Left Max Pathway Leakage 18,561 sccm May 2013 As-Found Min Pathway Leakage 9,838 sccm As-Left Max Pathway Leakage 20,883 sccm Unit 2 November 2010 As-Found Min Pathway Leakage 35,845 sccm As-Left Max Pathway Leakage 20,219 sccm April 2012 As-Found Min Pathway Leakage >La sccm As-Left Max Pathway Leakage 19,785 sccm November 2013 As-Found Min Pathway Leakage 11,193 sccm As-Left Max Pathway Leakage 18,800 sccm to AEP-NRC-2014-09 Page 11 Each unit has 120 mechanical penetrations and 55 electrical penetrations that are LLRT (Type B or C). Currently, approximately 5% of the components (Unit 1 and Unit 2 combined) require testing at an increased frequency due to leakage performance. However, neither unit's overall Type B and C leakage has approached the La leakage limit with the exception of the failure during U2C20 LLRT testing as noted above.

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 adopts the guidance in NEI 94-01, Revision 3-A, in place of NEI 94-01, Revision 0, for the Type C test interval (up to 75 months), but otherwise does not affect the scope or performance 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.

4.4 Supplemental Inspection Requirements Prior to initiating a Type A test, a general visual examination of accessible interior and exterior surfaces of the containment system is performed to identify any potential structural problems that could affect either the containment structure leakage integrity or the performance of the Type A test. This inspection is typically conducted in accordance with CNP's Containment Inservice Inspection (ISI) Plan, which implements the requirements of ASME, Section Xl, Subsection IWE/IWL. The applicable code edition and addenda for the second ten-year interval IWE/IWL program is the 2004 Edition.

The examinations performed in accordance with the IWE/IWL program satisfy the general visual examination requirements specified in 10 CFR 50, Appendix J, Option B. Identification and evaluation of inaccessible areas are addressed in accordance with the requirements of 10 CFR 50.55a(b)(2)(ix)(A) and (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). Each ten-year ISI interval is divided into three approximately equal-duration inspection periods. A minimum of one inspection during each inspection period of the ISI interval is required by the IWE/IWL program.

Subsection IWE assures that at least three general visual examinations of metallic components will be conducted before the next Type A test if the Type A test interval is extended to 15 years.

This meets the requirements of Section 9.2.3.2 of NEI 94-01, Revision 3-A and Condition 2 in Section 4.1 of the NRC SE for NEI 94-01, Revision 2.

Visual examinations of accessible concrete containment components in accordance with ASME Code, Section Xl, Subsection IWL are performed every five years, resulting in at least two IWL examinations being performed during a 15-year Type A test interval. CNP will perform additional exterior inspections as required per Appendix J when the Type A test frequency is extended to 15 years.

In addition to the IWL examinations, I&M performs a visual inspection of the accessible interior and exterior of the CNP Unit 1 and 2 Containment Buildings prior to each Type A test. This examination is performed in sufficient detail to identify any evidence of deterioration which may affect the reactor building's structural integrity or leak tightness. The examination is conducted to AEP-NRC-2014-09 Page 12 in accordance with approved plant procedures to satisfy the requirements of the 10 CFR 50 Appendix J Testing Program. The activity is coordinated with the IWE/IWL examinations to the extent possible.

Together these examinations assure that at least three general visual examinations of the accessible containment surfaces (exterior and interior) and one visual examination immediately prior to a Type A test will be conducted before the next Type A test if the Type A test interval is extended to 15 years, thereby meeting the requirements of Section 9.2.3.2 of NEI 94-01, Revision 3-A, as well as Condition 2 in Section 4.1 of the NRC SE for NEI 94-01, Revision 2.

The tables below provide dates of completed and scheduled ILRTs, completed containment surface examinations, along with an approximate schedule for future containment surface examinations, assuming the Type A test frequency is extended to 15 years.

Unit 1 Calendar Year Type A Test General Visual Examination of General Visual Examination of (ILRT) Accessible Exterior Surface Accessible Interior (Liner) Surface 2006 November 2006 August 2007 (IWL) October 2006 (IWE) 2007 2008 2009 August 2009 (Appendix J) October 2009 (IWE) 2010 2011 August 2011 (IWL) October 2011 (IWE) 2012 2013 2014 October 2014 IWE) 2015 2016 August 2016 (IWL) 2017 2018 2019 April 2019 (IWE) 2020 October 2020 October 2020 (Appendix J) October 2020 (IWE) 2021 August 2021 (IWL)

Unit 2 Calendar Year Type A Test General Visual Examination of General Visual Examination of (ILRT) Accessible Exterior Surface Accessible Interior (Liner) Surface 2006 April 2006 April 2006 (IWE) 2007 August 2007 (IWL) 2008 2009 August 2009 (Appendix J) April 2009 (IWE) 2010 2011 August 2011 (IWL) 2012 April 2012 (IWE) 2013 2014 2015 April 2015 (IWE) 2016 August 2016 (IWL) 2017 2018 2019 October 2019 (IWE) to AEP-NRC-2014-09 Page 13 2020 2021 April 2021 April 2021 (Appendix J) / August April 2021 (IWE) 2021 (IWL) 4.4.1 IWE Examinations A review was conducted for CNP Units 1 and 2 per IWE-1241, Examination Surface Areas (1992 Edition with 1992 Addenda of ASME Section Xl) for the initial 10-year Category E-C examination requirements. No areas were deemed susceptible to accelerated degradation and aging; therefore, augmented examinations per Category E-C were not required. This information is documented in the first 10-year Containment ISI Plan for CNP Units 1 and 2. The examinations performed during the first ten-year interval identified the following:

In 1998, a visual examination of the Unit 1 containment liner in the floor slab to liner area was performed after seal material removal to address the issue of liner corrosion as described in NRC Information Notice 97-10. Corrosion and pitting of the steel liner plate were identified along the moisture barrier seal near the containment cylinder base at elevation 598 ft., 9-3/8 in. Similar conditions were found in Unit 2. Based on a detailed engineering analysis, I&M concluded that the structural integrity of the containments to withstand normal operating, design basis accident, and severe accident loads was not affected by the as-found condition of the liner. The leak tight integrity of the containments was not impaired, and the liners, as-found, would have fulfilled their function as effective leak-tight barriers. This condition was documented in an I&M letter to the NRC dated March 3, 2000, (

Reference:

AEP:NRC:2612). . Modifications were made to the floor-liner seal to prevent further degradation of the liner. This issue was included in a special inspection pursuant to NRC Manual Chapter 0350, "Staff Guidelines for Restart Approval." Closure of the issue was documented in an NRC Inspection Report dated January 19, 2000, (

Reference:

AEP:NRC:2612).

The area behind the moisture barrier seal is normally inaccessible. In accordance with IWE-1220(b), the area is exempt from the examination requirements of IWE-2000, including the additional examination requirements of IWE-2430. The corrective action for the liner corrosion included modifying the seal design to prevent moisture intrusion and re-coating the affected area. Therefore, augmented examination per IWE-1240 is not required because the area is no longer likely to experience accelerated degradation or aging. Additionally, IWE-1240 does not apply to inaccessible areas. Although not required by IWE-2430 or IWE-1240, I&M has performed supplemental inspections of portions of the area behind the moisture barrier seal to verify the effectiveness of the new design. Sections of the redesigned moisture barrier seal were removed in both Unit 1 and Unit 2 approximately three years after installation, and a visual examination was performed on the liner area where corrosion was previously identified. The visual examination found no moisture intrusion and no active corrosion. Continued monitoring of the moisture barrier seal is performed with the scheduled VT-3 visual examination of the moisture barrier seal area each inspection period (3 inspections are required in a 10-year interval).

• In 1999, a visual examination of the containment liner identified an apparent weld repair of the liner plate. Surface preparation to allow further inspection dislodged repair to AEP-NRC-2014-09 Page 14 material exposing a hole through the liner plate. The hole was circular in appearance with a diameter of approximately 3/16 in. on the exterior surface and 3/4 in. on the interior surface. It appeared that the liner hole resulted from an inadequate repair of a hole drilled in error during plant construction. After the damaged liner plate section was cut out, a piece of wood, determined to be the handle of a wire brush, was found embedded in the concrete. Some minor corrosion was noted on the concrete side of the liner plate in the area of the embedded wire brush. This condition was also determined to be construction-related. The affected area of containment was restored to an acceptable design configuration. The repair was vacuum box tested and subjected to an LLRT. This condition was initially reported pursuant to 10 CFR 50.73 in Licensee Event Report (LER) 2000-001-00 (

Reference:

AEP:NRC:2612). Additional evaluation of the condition determined that the containment structure would have performed its safety related function during a design basis accident in the as-found condition, and the LER was retracted by LER 2000-001-01, (

Reference:

AEP:NRC:2612).

The through-wall hole in the liner plate did not invoke the requirements for augmented examination per IWE-1240 since the hole was not the result of degradation or aging.

The corrective action for the through-wall hole included removal of the wire brush and wooden handle to the maximum extent practical without cutting any stiffener steel, performing a repair of the concrete, and replacing the liner section that contained the hole. Therefore, augmented examination per IWE-1240 is not required because the area is no longer likely to experience accelerated degradation or aging since there is no wood in contact with the liner and the inadequate repair/hole has been eliminated.

In the second 10-year interval, which commenced March 1, 2010, one examination area was identified as being susceptible to accelerated degradation per IWE-1241 (2004 Edition of ASME Section Xl). A glycol pipe penetration in Unit 1 had a large amount of wet discoloration due to condensation below the insulated piping. After insulation removal, a VT-1 (detailed) visual examination was performed which revealed minor pitting that did not impact the leak tightness or structural integrity of the containment boundary. This area was classified as an Augmented Examination per Category E-C to get insulation removed for continued monitoring with a VT-1 visual examination during successive inspection periods.

CNP Units 1 and 2 have completed requirements of the first period, second 10-year Containment IWE ISI Program. The second 10-year interval IWE examination requirements use the 2004 Edition of ASME Section Xl as modified by the 10 CFR 50.55a(b) limitations for both units. Other than the augmented E-C examination discussed above, the remaining examinations are based on Category E-A and are visual (General, VT-3, and VT-1) examinations based on ASME Code or 10 CFR rule requirements.

In accordance with the Containment Inservice Testing Program, station personnel perform an IWE - General Visual examination on the accessible surface areas associated with the containment liner. Most of the conditions noted during the second interval IWE examinations consist of some flaking and discoloration of coatings along with some surface corrosion and minor pitting which did not impact the leak tightness or structural integrity of the containment boundary.

to AEP-NRC-2014-09 Page 15 4.4.2 IWL Examinations The second 10-year concrete containment examinations (IWL) have specified dates of August 24, 2011 and August 24, 2016, for Units 1 and 2. General and detailed visual examinations have been and will be completed in accordance with Category L-A of the code no earlier than or later than one year of the specified date for both units.

CNP Units 1 and 2 have completed or are completing the requirements of their second 10-year Containment ISI Program. Concrete containment examinations (IWL) were completed for Units 1 and 2 by the required date of August 24, 2011, in accordance with the requirements of the 2004 Edition of ASME Section Xl in the second 10-year interval. These examinations on the concrete exterior were conducted under the direction of the Responsible Engineer using the visual (VT-3C and VT-1C) method. The conditions noted during the IWL Examination included cracking, efflorescence, popouts, spalling, exposed rebar, and embedded foreign material.

Examination of the Unit 1 and Unit 2 containment structures did not reveal any significant observations that can potentially affect the structural integrity or the calculated design safety margins. Code repairs of the concrete have been performed for conditions such as exposed rebar, popout, and previous repair areas. Other areas that were not repaired will continue to be monitored in future examinations.

CNP Units 1 and 2 do not have an un-bonded post-tensioning system. As such, examinations required by Category L-B do not apply.

4.5 Deficiencies Identified Consistent with the guidance provided in NEI 94-01, Revision 3, Section 9.2.3.3, abnormal degradation of the primary containment structure identified during the conduct of IWE/IWL program examinations or at any other times is entered into the corrective action program for evaluation to determine the cause of the degradation and to initiate appropriate corrective actions.

4.6 Plant Specific Confirmatory Analysis 4.6.1 Methodology An evaluation has been performed assessing the risk impact of extending the CNP ILRT surveillance intervals from the current 10 years to 15 years. The evaluation is included as . This plant-specific risk assessment followed the guidance in NEI 94-01, Revision 3-A, the methodology described in EPRI TR-1018243, Revision 2-A of 1009325, and the NRC regulatory guidance outlined in RG 1.174 on the use of Probabilistic Risk Assessment (PRA) findings and risk insights in support of a request to change the licensing basis of the plant. In addition, the methodology used for Calvert Cliffs Nuclear Power Plant to estimate the likelihood and risk implication of corrosion-induced leakage of steel containment liners going undetected during the extended ILRT surveillance interval was also used for a sensitivity analysis.

The current CNP Level 1, internal events PRA (IEPRA) model results were used with a new Level 2 model to perform the plant-specific risk assessment. The Level 1 PRA model was peer reviewed against the ASME PRA Standard RA-S-2003. A gap assessment between the Level 1 to AEP-NRC-2014-09 Page 16 PRA model and current PRA Standard RA-Sa-2009 is addressed as a part of the PRA technical adequacy evaluation discussed in Enclosure 6. The Level 2 model was peer reviewed in 2013 against ASME/ANS PRA Standard RA-Sa-2009 with clarification and qualifications by the NRC in Regulatory Guide 1.200, Revision 2. The PRA analyses includes evaluations for the dominant external events (internal fire and seismic only, as high winds and tornados were considered negligible) from the CNP Individual Plant Examination of External Events. An entirely new fire PRA model was completed in 2013 and used in preparing the risk assessment associated with the amendment request.

In the SE issued by NRC letter dated June 25, 2008, the NRC concluded that the methodology in EPRI Report No. 1018243, Revision 2-A of 1009325, is acceptable for referencing by licensees proposing to amend their TS to extend the ILRT surveillance interval to 15 years, subject to the limitations and conditions noted in Section 4.2 of the SE. The following table addresses each of the four limitations and conditions for use of EPRI TR-1009325, Revision 2-A. These limitations and conditions were incorporated into EPRI TR-1018243.

From Safety Evaluation (SE) of EPRI - CNP Response 1009325

1. The licensee submits documentation CNP PRA technical adequacy is addressed in Section indicating that the technical adequacy of their 4.6.2.

PRA is consistent with the requirements of RG 1.200 relevant to the ILRT extension.

2. The licensee submits documentation EPRI Report No. 1018243, Revision 2-A of 1009325, indicating that the estimated risk increase incorporates these population dose and CCFP associated with permanently extending the acceptance guidelines, and these guidelines have been ILRT surveillance interval to 15 years is small used for the CNP assessment.

and consistent with the specific risk clarification provided in Section 3.2.4.5 of the SE. Specifically, a small increase in population dose should be defined as an increase in population dose of less than or equal to either 1.0 person-rem per year or 1 %

of the total population dose, whichever is less restrictive. In addition, a small increase in conditional containment failure probability (CCFP) should be defined as a value marginally greater than that accepted in a previous one-time ILRT extension requests.

This would require that the increase in CCFP be less than or equal to 1.5 percentage point.

3. The methodology in EPRI Report No. EPRI Report No. 1018243, Revision 2-A of 1009325, 1009325, Revision 2, is acceptable except for incorporated the use of 100 La as the average leak rate the calculation of the increase in expected for the pre-existing containment large leakage rate population dose (per year of reactor accident case (accident case 3b), and this value has operation). In order to make the methodology been used in the CNP specific risk assessment.

acceptable, the average leak rate accident (accident case 3b) used by the licensees shall be 100 La instead of 35 La.

to AEP-NRC-2014-09 Page 17

4. A LAR is required in instances where CNP Units 1 and 2 do not rely on containment containment over-pressure is relied upon for overpressure to assure adequate ECCS pump net emergency core cooling system (ECCS) positive suction head following design basis accidents.

performance. No additional risk analysis was needed for this consideration.

4.6.2 PRA Technical Adequacy CNP's Level 1 and large early release frequency (LERF) PRA model is characteristic of the as-built plant, and was used in the recent effort to transition the plant from Appendix R to National Fire Protection Association (NFPA)-805 license requirements. The current internal events model of record (09MORW) is a linked fault tree model. Severe accident sequences have been developed from internally initiated events. The sequences have been mapped to the radiological release end state (i.e., source term release to environment). CNP's PRA is based on a detailed model of the plant developed from the Individual Plant Examination which underwent NRC review. Review comments, current plant design, current procedures, plant operating data, current industry PRA techniques, and general improvements have been incorporated into the current PRA model. The model is maintained in accordance with CNP's PRA procedure.

CNP's PRA model has received the following independent, third-party reviews:

2001 Westinghouse Owners Group Peer Review - A full-scope Peer Review performed to the guidance of NEI 00-02. All of the Facts and Observations (F&Os) from that Peer Review that were graded at the A or B level were resolved. Only some of the F&Os that were graded at the C or D level were resolved.

2004 AREVA Gap Analysis - This effort identified differences between the 2003 ASME PRA Standard and the version used for the 2001 Peer Review.

2009 Westinghouse Owners Group Focused-Scope Peer Review - This limited-scope Peer Review addressed Supporting Requirements (SRs) that pertained to Common-Cause Failure modeling, data analysis performed in support of a Mitigating Systems Performance Indicator, Supplemental Diesel Generator system model and incorporation into Loss of Offsite Power/Station Blackout modeling, and Human Reliability Analysis for Steam Generator Tube Rupture response actions.

2012 Scientech performed a Gap Analysis in August of 2012 associated with CNP's LAR to transition to NFPA-805 to establish the quality of Cooks' IEPRA against the requirements of RG 1.200, Revision 2 by identifying differences between the SRs of the 2003 and 2009 Standards and the impact of these differences on the results of the previous reviews.

The 2012 gap assessment found that Test and Maintenance (T&M) factors were applied differently in the model than standard industry practice. The focused scope peer review associated with this gap has been finalized, and required revising the PRA model to conform to to AEP-NRC-2014-09 Page 18 industry practice for T&M treatment. The metrics for the ILRT extension reflect this recent PRA model change.

The 2012 gap assessment also identified several shortcomings of the LERF modeling which were applicable to the underlying Level 2 modeling. As a result, a Level 2 model update was conducted to address the modeling gaps identified. A focused-scope peer review was performed on the Level 2 model in 2013. The high-level results of this peer review are that 37 of the 41 LERF Analysis supporting requirements are met, while four of the supporting requirements are not met. The four Not Met supporting requirements are related to documentation issues and would have no numerical effect on the Level 2 model results. The LERF-to-core damage frequency (CDF) ratio estimate based on the revised Level 2 model is about 20% higher than the LERF-to-CDF ratio estimate based on the updated PRA model (i.e., 09MORW + T&M corrections).

CNP's updated PRA model, including the updated Level 2 model, is considered acceptable for use in assessing the risk impact of extending the CNP Units 1 and 2 containment ILRT surveillance interval to 15 years. Discussion of the various PRA model review results and possible impacts on PRA model technical adequacy is included in Enclosure 6.

4.6.3 Conclusion of Plant-Specific Risk Assessment Results The CNP risk assessment associated with extending the ILRT surveillance interval from three Type-A ILRT tests in 10 years to one Type-A ILRT test in 15 years is small, consistent with other generalized and specific industry analysis on this same subject. Details of the CNP risk assessment are contained in Enclosure 3. The CNP-specific results for extending the ILRT surveillance interval from the current 10 years to 15 years are summarized below.

1. The increase in LERF based on consideration of internal events only is conservatively estimated as 5.08E-08/yr. The guidance in RG 1.174 defines very small changes in LERF as those that are less than 1.OE-7/yr. Therefore, the estimated change in LERF is determined to be very small using the RG 1.174 guidelines. An assessment of the impact from external events (seismic and fire) was also performed. In this case, the total increase in LERF for combined internal and external events was conservatively estimated as 6.03E-07/yr for CNP Unit 1, and 5.62E-07/yr for CNP Unit 2. The total increase in LERF for the combined internal and external events model is determined to be "small" using RG 1.174 guidelines.

RG 1.174 also requires that when the calculated increase in LERF is in the range of 1.OE-07 per reactor year to 1.OE-06 per reactor year, applications will be considered only if it can be reasonably shown that the total LERF is less than 1.OE-05 per reactor year.

CNP's risk assessment in Enclosure 3 determined total LERF for the 15-year interval extension can be estimated as 9.60E-06/yr, below the RG 1.174 acceptance criteria for total LERF of 1.OE-05.

2. The calculated increase in population dose for all sequences is 6.92E-03 person-rem per year, which is 0.02% of total population dose determined in accordance with EPRI TR-1018243. EPRI TR-1018243 states that a small increase in population dose is defined as an increase of less than or equal to either 1.0 person-rem per year or 1 % of to AEP-NRC-2014-09 Page 19 the total population dose, whichever is less restrictive. Thus, the calculated population dose increase is small using the guidelines of EPRI TR-1018243. Moreover, the risk impact when compared to other severe accident risks is negligible.

3. The calculated increase in the CCFP is 0.32%. EPRI TR-1018243 states that an increase in CCFP of less than or equal to 1.5 percentage points is very small.

Therefore, the calculated CCFP increase meets the criteria of a very small increase.

4. As noted in the table preceding this section, CNP ECCS/engineering safeguards feature (ESF) pumps design basis do not credit containment overpressure to satisfy net positive suction head requirements for containment spray and low-head safety injection in recirculation mode during LOCAs. Thus, there is no change in CDF (or LERF) associated with the increase in the ILRT surveillance interval due to ECCS/ESF pump recirculation needs, meeting acceptance guidelines in RG 1.174 for a "very small" change in CDF and the impact of the Type-A ILRT extension is bounded by the LERF analysis.

5.0 REGULATORY ASSESSMENT 5.1 Applicable Regulatory Requirements/Criteria The propQsed 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 reviews "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.

NEI 94-01, Revision 3-A, describes an approach for implementing the performance-based requirements of 10 CFR 50, Appendix J, Option B. The document incorporates the regulatory positions stated in RG 1.163 and includes provisions for extending Type A and Type C intervals to 15 years and 75 months, respectively. NEI 94-01, Revision 3-A, delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate test frequencies. In the SEs issued by NRC letters dated June 25, 2008 and June 8, 2012, the NRC concluded that NEI 94-01, Revision 3-A, describes an acceptable approach for implementing the optional performance-based requirements of 10 CFR 50, Appendix J, and is acceptable for to AEP-NRC-2014-09 Page 20 referencing by licensees proposing to amend their TS with regard to containment leakage rate testing, subject to the limitations and conditions noted in Section 4.0 of the SEs.

EPRI TR-1009325, Revision 2, provides a risk impact assessment for optimized ILRT intervals up to 15 years, utilizing current industry performance data and risk informed guidance.

NEI 94-01, Revision 3-A, states that a plant-specific risk impact assessment should be performed using the approach and methodology described in TR-1009325, Revision 2, for a proposed extension of the ILRT interval to 15 years. In the SE issued by NRC letter June 25, 2008, the NRC concluded that the methodology in EPRI TR-1009325, Revision 2, is acceptable for reference by licensees proposing to amend their TS to extend the ILRT surveillance interval to 15 years, subject to the limitations and conditions noted in Section 4.0 of that SE.

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.

In conclusion, CNP 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.

5.2 No Significant Hazards Consideration A change is proposed to the CNP Units 1 and 2 TS 5.5.14, "Containment Leakage Rate Testing Program." The proposed amendment would replace the reference to RG 1.163 with a reference to NEI 94-01, Revision 3-A, issued July 2012, as the implementation document used by I&M to develop the CNP performance-based primary containment leakage testing program in accordance with Option B of 10 CFR 50, Appendix J. The proposed amendment would also extend the interval for the primary containment integrated leak rate test (ILRT), which is required to be performed by 10 CFR 50, Appendix J, from ten years to no longer than 15 years from the last ILRT and permit Type C testing to be performed at an interval of up to 75 months.

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

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

Response: No.

The proposed revision to TS 5.5.14 changes the testing period to a permanent 15-year interval for Type A testing (10 CFR Part 50, Appendix J, Option B, ILRT). The current test interval of 10 years would be extended to 15 years from the last Type A test. The proposed extension to Type A testing does not involve a significant increase in the consequences of an accident since research documented in NUREG-1493, "Performance-Based Containment System Leakage Testing Requirements," September 1995, has found that, generically, very few potential to AEP-NRC-2014-09 Page 21 containment leakage paths are not identified by Type B and C tests. NUREG-1493 concluded that reducing the Type A testing frequency to one per twenty years was found to lead to an imperceptible increase in risk. A high degree of assurance is provided through testing and inspection that the containment will not degrade in a manner detectable only by Type A testing.

The last Type A test (November 2006) shows leakage to be below acceptance criteria, indicating a very leak tight containment. Inspections required by the ASME Code Section Xl (Subsections IWE and IWL) and Maintenance Rule monitoring (10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants) are performed in order to identify indications of containment degradation that could affect that leak tightness.

Types B and C testing required by TSs will identify any containment opening such as valves that would otherwise be detected by the Type A tests. These factors show that a Type A test interval extension will not represent a significant increase in the consequences of an accident.

The proposed amendment involves changes to the CNP Units 1 and 2 10 CFR 50 Appendix J Testing Program Plan. The proposed amendment does not involve a physical change to the plant or a change in the manner in which the units are operated or controlled. The primary containment function is to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment itself 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, and do not involve any accident precursors or initiators. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased by the proposed amendment.

The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, for development of the CNP performance-based leakage testing program. Implementation of these guidelines continues to provide adequate assurance that during design basis accidents, the primary containment and its components will limit leakage rates to less than the values assumed in the plant safety analyses. The potential consequences of extending the ILRT interval from 10 years to 15 years have been evaluated by analyzing the resulting changes in risk. The increase in risk in terms of person-rem per year resulting from design basis accidents was estimated to be acceptably small, and the increase in the LERF resulting from the proposed change was determined to be within the guidelines published in NRC RG 1.174. Additionally, the proposed change maintains defense-in-depth by preserving a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation.

I&M has determined that the increase in CCFP due to the proposed change would be very small.

Therefore, it is concluded that the proposed amendment does not significantly increase the consequences of an accident previously evaluated.

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

Response: No.

The proposed revision to TS 5.5.14 changes the testing period to a permanent 15-year interval for Type A testing (10 CFR Part 50, Appendix J, Option B, ILRT. The current test interval of 10 years, based on past performance, would be extended to 15 years from the last Type A test to AEP-NRC-2014-09 Page 22 (November 2006). The proposed extension to Type A testing does not create the possibility of a new or different type of accident since there are no physical changes being made to the plant and there are no changes to the operation of the plant that could introduce a new failure mode creating an accident or affecting the mitigation of an accident.

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

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

Response: No.

The proposed revision to TS 5.5.14 changes the testing period to a permanent 15-year interval for Type A testing (10 CFR Part 50, Appendix J, Option B, ILRT. The current test interval of 10 years, based on past performance, would be extended to 15 years from the last Type A test (November 2006). The proposed extension to Type A testing will not significantly reduce the margin of safety. NUREG-1493, "Performance-Based Containment System Leakage Testing Requirements," September 1995, generic study of the effects of extending containment leakage testing, found that a 20 year extension to Type A leakage testing resulted in an imperceptible increase in risk to the public. NUREG-1493 found that, generically, the design containment leakage rate contributes about 0.1% to the individual risk and that the decrease in Type A testing frequency would have a minimal effect on this risk since 95% of the potential leakage paths are detected by Type C testing. Regular inspections required by the ASME Code Section Xl (Subsections IWE and IWL) and maintenance rule monitoring (10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants) will further reduce the risk of a containment leakage path going undetected.

The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 3-A, for development of the CNP performance-based leakage testing program, and establishes a 15-year interval for the performance of the primary containment ILRT. The amendment does not alter the manner in which safety limits, limiting safety system setpoints, or limiting conditions for operation are determined. The specific requirements and conditions of the 10 CFR Part 50, Appendix J Testing Program Plan, as defined in the TS, ensure that the degree of primary containment structural integrity and leak-tightness that is considered in the plant safety analyses is maintained. The overall containment leakage rate limit specified by the TS is maintained, and the Type A, B, and C containment leakage tests will continue to be performed at the frequencies established in accordance with the NRC-accepted guidelines of NEI 94-01, Revision 3-A.

Containment inspections performed in accordance with other plant programs serve to provide a high degree of assurance that the containment will not degrade in a manner that is detectable only by an ILRT. In addition, CNP has a containment monitoring capability for the detection of gross containment leakage that may develop during power operation. This combination of factors ensures that evidence of containment structural degradation is identified in a timely manner. Furthermore, a risk assessment using the current CNP PRA model concluded that extending the ILRT test interval from 10 years to 15 years results in a very small change to the CNP risk profile.

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

to AEP-NRC-2014-09 Page 23 In conclusion, based on the considerations discussed 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 be conducted in compliance with the NRC's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. I&M 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.

5.3 ENVIRONMENTAL CONSIDERATION

I&M has evaluated this LAR against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. I&M has determined that this LAR meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR Part 50 that changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, or that changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria:

(i) The amendment involves no significant hazards consideration.

As demonstrated in Section 4.3, the proposed TS change does not involve a significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

This LAR will not change the types or amounts of any effluents that may be released offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

This LAR will not increase the individual or cumulative occupational radiation exposure.

6.0 CONCLUSION

NEI 94-01, Revision 3-A, describes an NRC-accepted approach for implementing the performance-based requirements of 10 CFR 50, Appendix J, Option B. It incorporates the regulatory positions stated in RG 1.163 and includes provisions for extending Type A and Type C test intervals to 15 years and 75 months, respectively. NEI 94-01, Revision 3-A delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance test frequencies. I&M is adopting the guidance of NEI 94-01, Revision 3-A for the CNP Units 1 and 2 10 CFR 50, Appendix J testing program plan.

Based on the previous ILRT tests conducted at CNP Units 1 and 2, it may be concluded that extension of the containment ILRT surveillance interval from 10 to 15 years represents minimal to AEP-NRC-2014-09 Page 24 risk to increased leakage. The risk is minimized by continued Type B and Type C testing performed in accordance with 10 CFR 50, Appendix J, Option B and inspection activities performed as part of the CNP IWE/IWL ASME Section Xl ISI program.

This experience is supplemented by risk analysis studies, including the CNP risk analysis in to this letter. The findings of the CNP risk assessment confirm the general findings of previous studies, on a plant-specific basis, that extending the ILRT surveillance interval from 10 to 15 years results in a minimal change to the CNP Units 1 and 2 risk profile.

7.0 PRECEDENCE This request is similar to the following license amendments, which have been approved by the NRC:

1. Nine Mile Point Nuclear Station, Unit 2 - Issuance of Amendment Re: Extension of Primary Containment Integrated Leakage Rate Testing Interval (TAC No. ME1650, ADAMS Accession Number ML100730032) approved March 30, 2010.

2. Arkansas Nuclear One, Unit No.2 - Issuance of Amendment Re: Technical Specification Change to Extend the Type A Test Frequency to 15 Years (TAC No. ME4090, ADAMS Accession Number ML110800034) approved April 7, 2011.

3. Palisades Nuclear Plant - Issuance of Amendment to Extend the Containment Type A Leak Rate Test Frequency to 15 Years (TAC No. ME5997, ADAMS Accession Number ML120740081) approved April 23, 2012.

8.0 REFERENCES

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

2. NEI document NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J."

3. NEI document NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 0, dated July 21, 1995.

4. EPRI report TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals," dated August 1994.

5. NEI document NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 2, dated August 2007. (ADAMS Accession Number ML072970206).

6. Letter from M. J. Maxin, NRC, to J. C. Butler, NEI, "Final Safety Evaluation For Nuclear Energy Institute (NEI) Topical Report (TR) 94-01, Revision 2, "Industry Guideline For to AEP-NRC-2014-09 Page 25 Implementing Performance-Based Option of 10 CFR PART 50, Appendix J" and Electronic Power Research Institute (EPRI) Report No. 1009325, Revision 2, August 2007, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals" (TAC NO. MC9663), dated June 25, 2008.

7. EPRI report TR-1018243, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325", dated October 2008.

8. Letter from Sher Bahadur, NRC, to Mr. Biff Bradley, NEI, "Final Safety Evaluation of Nuclear Energy Institute (NEI) Report, 94-01, Revision 3, "Industry Guideline For Implementing Performance-Based Option of 10 CFR PART 50, Appendix J" (TAC NO.

ME2164), dated June 08, 2012.

9. NRC Regulatory Issue Summary 2008-27, Staff Position on Extension of the Containment Type A Test Interval Beyond 15 Years Under Option B of Appendix J to 10 CFR Part 50, dated December 8, 2008.