AEP-NRC-2020-72, License Amendment Request for One-Time Extension of the Containment Type a Leak Rate Testing Frequency

License Amendment Request for One-Time Extension of the Containment Type a Leak Rate Testing Frequency

ML20363A011
Person / Time
Site:	Cook
Issue date:	12/14/2020
From:	Lies Q Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2020-72
Download: ML20363A011 (67)

Text

Indiana Michigan Power Cook Nuclear P1Blll One Cook Place Bndgman, Ml 49106 1ndianamrchiganpower com An~Coo¥JB'l)'

BOUNDLESS ENERGY-December 14, 2020 AEP-NRC-2020-7 2 10 CFR 50.90 Docket No.: 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Donald C. Cook Nuclear Plant, Unit 2 LICENSE AMENDMENT REQUEST FOR ONE-TIME EXTENSION OF THE CONTAINMENT TYPE A LEAK RATE TESTING FREQUENCY Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 2, is submitting a request for an amendment to the Technical Specifications (TS) for CNP Unit 2. The proposed license amendment request would allow for a one-time extension to the 15-year frequency of the CNP Unit 2 containment leakage rate test (i.e., Integrated Leak Rate Test (ILRT) or Type A test). This test is required by TS 5.5.14, Containment Leakage Rate Testing Program. The proposed one-time change would pem1it the current ILRT interval of 15 years to be extended by approximately eighteen months to no later than the plant startup after the fall 2022 refueling outage.

The most recently completed Type A test at CNP Unit 2 was on April 22, 2006, therefore the next Type A test is due to start no later than April 30, 2021. In order to perfom1 the test during a regularly scheduled refueling outage, the next Type A test at CNP Unit 2 would need to be completed no later than startup after the spring 2021 refueling outage.

On January 31, 2020, the U.S. Department of Health and Human Services declared a public health emergency for the United States to aid the nation's healthcare community in responding to the Novel Coronavirus and its associated disease, COVID-19. On March 10, 2020, Michigan Governor Gretchen Whitmer declared a state of emergency. The COVID-19 outbreak was subsequently characterized as a pandemic by the World Health Organization on March 11, 2020, and on March 13, 2020, President Donald Trump declared the COVID-19 pandemic a national emergency.

On October 2, 2020, the U.S. Department of Health and Human Services Secretary Alex Azar announced the renewal of the COVID-19 national public health emergency declaration, effective October 23, 2020.

In response to concerns of a continuation of the COVID-19 public health emergency, in the interest of personnel safety, and to preclude the potential for transmittal and spread of COVID-19, l&M requests a one-time extension of the CNP Unit 2 Type A test interval. This request is part of an }bD /

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U. S. Nuclear Regulatory Commission AEP-NRC-2020-72 Page2 overall effort by l&M to reduce the number of outside personnel required on-site, and the overall outage scope, in response to the developing COVID-19 pandemic situation while maintaining the safety and reliability of the plant for the next operating cycle. This effort by l&M assures that the overriding priority of nuclear safety is maintained while providing for p)ant personnel and public safety and health. Performing the Type A test would require 25 vendor personnel from across the United States working alongside plant personnel in close proximity for extended periods of time. Including the ILRT in the spring outage scope would aJso increase the overall outage duration by approximately two days, increasing the amount of time that supplemental workforce would remain on-site.

Current scope reduction efforts have reduced the number of scheduled man-hours from an estimate of 240,000 to approximately 150,000 man-hours, and reduced the number of outside personnel required from an estimate of 1,250 to approximately 850 personnel. This reduction in scope and required outside personnel will allow CNP Unit 2 outage personnel to more effectively follow guidelines for social distancing established by the Centers for Disease Control and Prevention and Michigan Department of Health and Human Services during the spring outage.

Enclosure 1 provides an affirmation statement pertaining to the information contained herein.

Enclosure 2 provides a description and assessment of the proposed changes. Enclosure 3 provides an assessment of risk associated with the one-time extension. Enclosure 4 provides Unit 2 TS pages, marked to show the proposed changes. New clean TS pages with proposed changes incorporated will be provided to the Nuclear Regulatory Commission (NRC) Licensing Project Manager when requested.

In accordance with 10 CFR 50.91, a copy of this application, with enclosures, is being provided to the designated Michigan state officials. .,

l&M requests approval of the proposed amendment by March 4, 2021, in order to facilitate preparation and planning for the next Unit 2 refueling outage, currently scheduled to occur during spring 2021.

Once approved, the amendment shall be implemented within 30 days.

There are no regulatory commitmen'ts made in this submittal. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Director, at (269) 466-2649.

Sincerely, Q.t~l~

Site Vice President Indiana Michigan Power Company BMC/kmh

U. S. Nuclear Regulatory Commission AEP-NRC-2020-72 Page 3

Enclosures:

1. Affirmation

2. Description and Assessment of the Technical Specification Changes

3. Evaluation of Risk Significance of Short-Term ILRT Extension

4. Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages Marked to Show Proposed Changes c: , R. J. Ancona - MPSC EGLE - RMD/RPS J. 8. Giessner - NRC Region Ill D. L. Hille - AEP Ft. Wayne, w/o enclosures NRC Resident Inspector R. M. Sistevaris - AEP Ft. Wayne, w/o enclosures S. P. Wall - NRC Washington, D.C.

A J. Williamson - AEP Ft. Wayne, w/o enclosures

U. S. Nuclear Regulatory Commission AEP-NRC-2020-72 Page 4 be: S. A. Dailey K. J. Femeau J.P. Gebbie R. B. Haemer K. M. Harper H. L. Levendosky Q. S. Lies J.M. Petro R. Ramirez M. K. Scarpello M. D. Sartwell

Enclosure 1 to AEP-NRC-2020-72 AFFIRMATION I, Q. Shane Lies, being duly sworn, state that I am the Site Vice President of Indiana Michigan Power Company (l&M), that I am authorized to sign and file this request with the U. S. Nuclear Regulatory Commission on behalf of l&M, and that the statements made and the matters set forth herein pertaining to l&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS / l..( DAY OF Vecem,kr 2020 My Commission Expires o/)-;}O-plS

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Enclosure 2 to AEP-NRC-2020-72 Description and Assessment of Technical Specification Changes

1.0 DESCRIPTION

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 2, is submitting a request for an amendment to the Technical Specifications (TS) for CNP Unit 2. The proposed license amendment request (LAR) would allow for a one-time extension to the 15-year frequency of the CNP Unit 2 containment leakage rate test (i.e.,

Integrated Leak Rate Test (ILRT) or Type A test). This test is required by TS 5.5.14, Containment Leakage Rate Testing Program. The proposed one-time change would permit the current ILRT interval of 15 years to be extended by approximately eighteen months to no later than the plant startup after the fall 2022 refueling outage.

The most recently completed Type A test at CNP Unit 2 was on April 22, 2006, therefore the next Type A test is due to start no later than April 30, 2021. In order to perform the test during a regularly scheduled refueling outage, the next Type A test at CNP Unit 2 would need to be completed no later than startup after the spring 2021 refueling outage.

On January 31, 2020, the U.S. Department of Health and Human Services (DHHS) declared a public health emergency for the United States to aid the nation's healthcare community in responding to the Novel Coronavirus and its associated disease, COVID-19. On March 10, 2020, Michigan Governor Gretchen Whitmer declared a state of emergency. The COVID-19 outbreak was subsequently characterized as a pandemic by the World Health Organization on March 11, 2020, and on March 13, 2020, President Donald Trump declared the COVID-19 pandemic a national emergency.

On October 2, 2020, the U.S. DHHS Secretary Alex Azar announced the renewal of the COVID-19 national public health emergency declaration, effective October 23, 2020.

In response to concerns of a continuation of the COVID-19 public health emergency, in the interest of personnel safety, and to preclude the potential for transmittal and spread of COVID-19, l&M requests a one-time extension of the CNP Unit 2 Type A test interval. This request is part of an overall effort by l&M to reduce the number of outside personnel required on-site, and the overall outage scope, in response to the developing COVID-19 pandemic situation while maintaining the safety and reliability of the plant for the next operating cycle. This effort by l&M assures that the overriding priority of nuclear safety is maintained while providing for plant personnel and public safety and health. Performing the Type A test would require 25 vendor personnel from across the United States working alongside plant personnel in close proximity for extended periods of time. Including the ILRT in the spring outage scope would also increase the overall outage duration by approximately two days, increasing the amount of time that supplemental workforce would remain on-site.

Current scope reduction efforts have reduced the number of scheduled man-hours from an estimate of 240,000 to approximately 150,000 man-hours, and reduced the number of outside personnel required from an estimate of 1,250 to approximately 850 personnel. This reduction in scope and required outside personnel will allow CNP Unit 2 outage personnel to more effectively follow to AEP-NRC-2020-72 Page 2 guidelines for social distancing established by the Centers for Disease Control and Prevention and Michigan Department of Health and Human Services during the spring outage.

2.0 PROPOSED CHANGE

CNP Unit 2 TS 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(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," dated July 2012, and Section 4.1, "Limitations and Conditions for NEI TR 94-01, Revision 2," of the NRC Safety Evaluation Report in NEI 94-01, Revision 2-A, dated October 2008.

The proposed change to CNP Unit 2 TS 5.5.14, Containment Leakage Rate Testing Program, will add an exception to allow for the perfom1ance of the next Type A test no later than the fall 2022 refueling outage for Unit 2, as follows (added text in bold italic):

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Perfom1ance-Based Option of 10 CFR 50, Appendix J," dated July 2012, and Section 4.1, "Limitations and Conditions for NEI TR 94-01, Revision 2," of the NRC Safety Evaluation Report in NEI 94-01, Revision 2-A, dated October 2008, except that the next Type A test performed after the April 22, 2006, Type A test shall be performed no later than the plant startup after the 'fall 2022 refuellng outage.

Pages from the CNP Unit 2 TS, marked up to show the proposed changes, are included as Enclosure 4 to this letter. The additional language proposed by this amendment request is shown as boxed text.

3.0 BACKGROUND

The CNP Unit 2 containment is a steel-lined, reinforced concrete structure. The containment structure, including all its penetrations, includes a low-leakage steel liner designed to contain the radioactive material that may be released from the reactor core following a design basis loss of coolant accident. Additionally, the containment structure provides shielding from the fission products that may be present in the containment atmosphere following accident conditions.

As required by 10 CFR 50.54(0), the CNP Unit 2 containment is subject to the requirements set forth in 10 CFR 50, Appendix J. The testing requirements of 10 CFR 50, Appendix J, provide assurance that leakage from containment, including systems and components that penetrate 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 perfom1ed so that proper maintenance and repairs are perfom1ed on the systems and components penetrating containment.

The limitation on containment leakage provides assurance that containment would perform its design function following an accident up to and including the plant design basis accident. Appendix J to AEP-NRC-2020-72 Page 3 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 (Cl) 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 testing interval but does not change the Appendix J Type A, Type B, or Type C test methods.

Chronology of 10 CFR 50 Appendix J Testing Requirements 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. 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, Regulatory Guide (RG) 1.163 (Reference 1) was issued. The RG endorsed NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J" (Reference 2), with certain modifications and additions. Option B, in concert with RG 1.163 and NEI 94-01, Revision 0, allowed licensees with a satisfactory ILRT performance history (i.e., two consecutive successful type A tests) to reduce the frequency of the ILRT from three tests in ten years to one test in ten years. This relaxation was based on an Nuclear Regulatory Commission (NRC) risk program and Electric Power Research Institute (EPRI)

Topical Report (TR)-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals" (Reference 3), which illustrated that the risk increase associated with extending the ILRT surveillance interval was very small.

By letter dated August 31, 2007, NEI submitted NEI 94-01, Revision 2 (Reference 4), to the NRC Staff for review. NEI 94-01, Revision 2, describes an approach for implementing the performance-based requirements of Option B, which includes provisions for extending Type A intervals to up to 15 years and incorporates the regulatory positions stated in RG 1.163 (Reference 1).

It delineates a performance-based approach for determining containment leakage rate surveillance testing frequencies using industry performance data, plant-specific performance data, and risk insights. The NRC final Safety Evaluation (SE) issued by letter dated June 25, 2008 (Reference 5),

documents the 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 was subsequently issued as Revision 2-A, dated October 2008 (Reference 6).

On December 8, 2008, the NRC issued Regulatory Issue Summary (RIS) 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" (Reference 7). The RIS clarifies the NRC position concerning licensee requests to extend Type A test intervals beyond 15 years, stating that a licensee can commence the test no later than the last day of the month in which it becomes due, without seeking NRC approval through a license amendment. The RIS also endorses the statement made in NEI 94-01, Revision 2, that if the test interval ends while primary containment integrity is not required, or is required solely for

Enclosure 2 to AEP-NRC-2020-72 Page 4 shutdown activities, the test interval may be extended indefinitely, but a Type A test shall be completed prior to entering the operating mode requiring primary containment integrity.

By letter dated June 9, 2011, NEI submitted NEI 94-01 Revision 3 (Reference 8) to the NRC Staff for review. NEI 94-01, Revision 3, added guidance for extending Type C Local Leak Rate Test (LLRT) surveillance intervals beyond sixty months. The NRC final SE issued by letter dated June 8, 2012 (Reference 9), 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 (Reference 10).

Current CNP Unit 2 Testing Requirements Under 10 CFR 50 Appendix J Option B By letter dated March 7, 2014, as $Upplemented by letters dated September 30, 2014, December 16, 2014, January 15, 2015, and February 20, 2015, l&M submitted an amendment request to allow a permanent extension of the Type A primary containment integrated leak rate test frequency from once every 1O years to once every 15 years. On March 30, 2015, the NRC approved Amendment No. 309 for CNP Unit 2, authorizing the adoption of NEI 94-01 Rev 3-A as the implementation document to develop the performance-based primary containment leakage testing program at CNP Unit 2, in accordance with 10 CFR Part 50, Appendix J, Option B, and allowing l&M to extend the containment Type A test interval for CNP Unit 2 from 10 years to 15 years.

With the approval of CNP Unit 2 Amendment 309 (Reference 11 ), the due date for the Unit 2 Type A test moved from April 22, 2016, to April 22, 2021. The proposed change would defer the Type A test for CNP Unit 2 until no later than the startup after the fall 2022 refueling outage, which is currently scheduled to begin on October 8, 2022. This represents an extension of approximately eighteen months. The intervals for the Type B and Type C tests at CNP Unit 2 would remain unchanged at 120 months and 75 months, respectively.

4.0 TECHNICAL ANALYSIS

As required by 10 CFR 50.54(0), the CNP Unit 2 containment is subject to the requirements set forth in 10 CFR 50, Appendix J. Option B of Appendix J requires that containment leakage test intervals be determined using a performance-based approach. In a letter dated March 30, 2015 (Reference 11 ), the NRC approved a license amendment submitted by l&M, extending the Type A testing interval to 15 years, and establishing April 22, 2021, as the date by which l&M is to conduct the next CNP Unit 2 Type A test.

The NRC clarified its position concerning licensee requests to extend Type A tests beyond the currently approved 15 years in RI$ 2008-27 (Reference 7), stating that any extension beyond the end of the month in which the test is due would require a license amendment request.

Per RIS 2008-27, the license amendment request should demonstrate:

• A sound technical justification and/or undue h.ardship or unusual difficulty

• The requested amendment poses minimal safety risk

• Acceptable plant-specific containment performance, including a plant-specific risk-informed analysis

• That containment does not have a history of significant degradation issues

Enclosure 2 to AEP-NRC-2020-72 Page 5 4.1 Description of Containment The CNP Unit 2 containment is a steel-lined, reinforced concrete structure. The containment structure, including all its penetrations, includes a low-leakage steel liner designed to contain. the radioactive material that may be released from the reactor core following a design basis loss of coolant accident. Additionally, the containment structure provides shielding from the fission products that may be present in the containment atmosphere following accident conditions. 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 tum 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 ACl-318-63 and ACl-301-66. The structural steel components were designed in accordance with the American Institute of Steel Construction, AISC-69 specifications.

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, to AEP-NRC-2020-72 Page 6 located circumferentially between the crane wall and the outer wall of the containmeat 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.

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 (°F) in lower containment, and between 60 and 100°F in upper containment.

4.2 Integrated Leak Rate History Previous CNP Unit 2 ILRT results have confirmed that the containment is acceptable, with considerable margin, with respect- to the TS acceptance criterion of 0.25 percent (%) leakage of containment air weight per day at the design basis loss of coolant accident pressure. Since the last three Type A test results meet the performance leakage rate criteria from NEI 94-01, Revision 3-A (Reference 10), a test frequency of 15 years would be acceptable.

Enclosure 2 to AEP-NRC-2020-72 Page 7 It should be noted that Amendment 314 to CNP Unit 2 TS, issued in October 2016 (Reference 12),

changed the value of the allowable leakage rate (La) from 0.25% of containment air weight per day to 0.18% of containment air weight per day. When implementing the TS change into the ILRT procedure and the calculation of La, a more conservative value for containment free volume was also used, resulting in a change of La from 110,219 standard cubic centimeters per minute (seem) to 68,559 seem in March of 2017. However, even comparing the past ILRT leakage to the newer, more stringent value of La shows significant margin.

Unit 2 ILRT Results (Type A Test)

Test Date Performance Criterion Acceptance Limit*

February 1989 Mass point Upper Confidence Limit 0.075 of La 1.0 La (UCL) leakage with penalties: 0.121 of new La**

May 1992 Mass point UCL leakage with penalties: 0.109 of La 1.0 La 0.175 of new La**

April 2006 Mass point UCL leakage with penalties: 0.234 of La 1.0 La 0.376 of new La**

• The total allowable "as-left" leakage is 0. 75 La

• • The new La, established in March 2017, is 68,559 seem, compared to the La In use at the time of the test, which was 110,219 seem No modifications that require a Type A test are planned at CNP Unit 2 prior to the fall 2022 refueling outage when the next Type A test will be performed in accordance with this proposed change. Any unplanned modifications to 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 Unit 2 containment which could have adversely affected containment integrity. There is no anticipated addition or removal of plant hardware within Unit 2 containment which could affect leak-tightness.

• 4.3 Type 8 and Type C Testing Programs CNP Unit 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 Band TS 5.5.14. The Type Band 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. As discussed in NUREG-1493 (Reference 13), Type Band Type C tests can identify the.vast majority (greater than 95%) of all potential containment leakage paths. The results of the test program are used to demonstrate that proper maintenance and repairs are made on these components throughout their service life.

A review of the most recent Type B and Type C test results and a comparison with the allowable leakage rate was performed. In one instance, in the spring of 2012, the failure of containment isolation check valve 2-Sl-189 caused LLRT results to exceed allowable leakage. During that outage, leakage

- from all other containment penetrations combined was found to be 11,286 seem, or roughly 10.2% of La. With the exception of the spring 2012 LLRT, the combined Type B and Type C leakage for CNP

Enclosure 2 to AEP-NRC-2020-72 Page 8 Unit 2 has consistently remained well below allowable leakage (0.6 La). The as-found minimum and as-left maximum pathway leak rate summary totals for the last several refueling outages are shown below.

Unit 2 LLRT Results* (Type Band Type C Tests)

As-Found As-Found As-Left As-Left La 0.6La Minimum Minimum as Maximum Maximum as (seem) (seem)

(seem)  % of 1.0La (seem)  % of 1.0La Fall 2010 110,219 66,131 35,845 32.52% 20,219 18.34%

Spring 2012 110 219 66,131 > La ...... > La 19,785 17.95%

Fall 2013 110 219 66,131 11,193 10.16% 18 800 17.06%

Spring 2015 110 219 66,131 16,635 15.09% 44,969 40 80%

Fall 2016 110,219 66,131 35,013 31.77% 15,544 14.10%

Spring 2018 68 559 41,135 9,856 14.38% 17,194 25.08%

Fall 2019 68,559 41,135 16,028 23.38% 18,502 26.99%

• the total allowable as-found minimum pathway leakage, or as-left maximum pathway leakage is 0.6La

...... leakage other than through check valve 2-Sl-189 was 11,286 seem 4.4 Supplemental Inspection Requirements In the SE for NEI 94-01, Revision 2 (Reference 5), the NRC stated the following requirement for the performance of Supplemental Visual Inspections in the SE Section 3.1.1.3, "Adequacy of Pre-Test Inspections (Visual Inspections):"

Subsections IWE and IWL of the ASME Code,Section XI, as incorporated by reference in 10 CFR 50.55a, require general visual examinations two times within a 10-year interval for concrete components (Subsection IWL), and three times within a 10-year interval for steel components (Subsection IWE). To avoid duplication or deletion of examinations, licensees using NEI TR 94-01, Revision 2, have to develop a schedule for containment inspections that satisfy the provisions of Section 9.2.3.2 of this TR and ASME Code,Section XI, Subsection IWE and IWL requirements.

The second ten-year containment inservice inspection (CISI) interval began March 1, 2010, and concluded February 29, 2020. The CISI Program Plan for the second interval was developed in accordance with the requirements of the 2004 Edition, No Addenda, of the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code,Section XI, Subsections IWE and IWL for Inspection Program B, as modified by 10 CFR 50.55a. Identification and evaluation of inaccessible areas during the second ten-year CISI interval are addressed in accordance with the requirements of 10 CFR 50.55a(b )(2)(ix)(A) for IWE and 10 CFR 50.55a(b )(2)(viii)(E) for IWL.

Examination of pressure-retaining bolted connections and evaluation of containment bolting flaws or degradation during the second ten-year CISI interval are performed in accordance with the requirements of 10 CFR 50.55a(b)(2)(ix)(G) and 10 CFR 50.55a(b)(2)(ix)(H).

The third ten-year CISI interval began March 1, 2020, and will conclude on February 28, 2030. The CISI Program Plan for the third interval is based on the rules of ASME Section XI, Subsections IWE and IWL, 2013 Edition, as modified by 10 CFR 50.55a. Identification and evaluation of inaccessible to AEP-NRC-2020-72 Page 9 areas during the third ten-year CISI interval are addressed in accordance with the requirements of 10 CFR 50.55a(b )(2)(ix)(A) for IWE and 10 CFR 50.55a(b )(2)(viii)(H) and (I) for IWL. Requirements for examination of pressure-retaining bolted connections and evaluation of containment bolting flaws or degradation located in 10 CFR 50.55a(b )(2)(ix)(G) and (H) have been incorporated into the 2013 Edition of Section XI applicable to the third ten-year CISI interval.

Each ten-year CISI interval is divided into three approximately equal-duration inspection periods. A minimum of one inspection during each inspection period of the CISI interval is required by the IWE program. Visual examinations of accessible concrete containment components in accordance with ASME Code,Section XI, Subsection IWL are performed every five years, resulting in at least two IWL examinations being performed during a 15-year Type A test interval. 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.

In addition to the IWL examinations, l&M performs a visual inspection of the accessible interior and exterior of the CNP Unit 2 Containment Building 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 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.

The table below provide dates of completed and scheduled ILRTs, completed containment surface examinations, along with an approximate schedule for future containment surface examinations.

Unit2 Type A Test General Visual Examination of General Visual Examination of Calendar Year (ILRT) Accessible Exterior Surface Accessible Interior (Liner) Surface 2006 April 2006 April 2006 (IWE) 2007 August 2007 (IWL) 2008 2009 August 2009 (Aooendix J) April 2009 (IWE) 2010 2011 August 2011 (IWL) 2012 April 2012 (IWE) 2013 2014 2015 April 2015 (IWE) 2016 2017 July 2017 (IWL) 2018 2019 October 2019 (IWE) 2020 2021 2022 October 2022 AuQust 2022 (IWL) October 2022 (IWE) 4.4.1 IWE Examinations A review of CNP Unit 2 containment surfaces was conducted when establishing CISI jurisdictional boundaries during the initial CISI Program, per IWE-1241, Examination Surface Areas (1992 Edition

Enclosure 2 to AEP-NRC-2020-72 Page 10 with 1992 Addenda of ASME Section XI), 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 CISI Plan for CNP Unit 2.

The April 2009, IWE examination documented that no conditions were identified that did not meet acceptance standards. The April 2012, and April 2015, IWE examinations found some areas of light rust and discoloration caused by condensation, as well as some delamination of topcoat and flaking or peeling paint. No areas of structural distress or wastage were identified. The most recent IWE examination occurred in October 2019, and reported that no conditions were identified that did not meet acceptance standards.

Since-the last ILRT a contair'fnierit interior-surface- coating inspection was performed each outage that included the liner plate as part of the Safety-Related Coatings program. Additionally, four IWE inspections have been completed on CNP Unit 2 since the last ILRT. Either the IWE inspections or the Safety-Related Coatings program inspections would satisfy the Appendix J interior inspection requirements and neither has indicated any degradation in the containment liner that would prevent it from fulfilling its leak-tight integrity purpose for 10 CFR Part 50, Appendix J. An additional IWE inspection will be performed during the fall 2022 refueling outage.

4.4.2 IWL Examinations Since the last ILRT, in April 2006, there have been three ASME Section XI, Subsection IWL examinations completed, with the most recent examination taking place in July of 2017. These examinations on the concrete exterior were conducted under the direction of the Responsible Engineer using the General and Detailed Visual Exam methods. The first IWL examination to be completed after the 2006 Type A test was performed in 2007, in accordance with the requirements of the 1992 Edition with the 1992 Addenda of ASME Section XI, as part of the first 10-year interval of the CISI Program. The actual IWL examination took place over the course of several months, between February 2006, and July 2007, overlapping the timeframe of the most recent Type A test.

The IWL examination completed in July of 2007 did not reveal any significant observations that could potentially affect the structural integrity of the Unit 2 containment or the calculated design safety margins. During the 2007 inspection, a small section (approximately one square inch) of spalled concrete was discovered in the Unit 2 containment exterior, at azimuth 270 degrees, elevation 695 ft., exposing either rebar or a mechanical rebar splice connector. The other recordable observations consisted of four surface cracks between 1/32 in, and 1/16 ir,., popout and spalling of maximum 1 in.

depth, and efflorescence and loosening of some previously installed patches on the Unit 2 dome.

The conditions observed in the 2007 inspection were evaluated and determined to not pose a threat to the design basis margin for the concrete containment structure. The concrete at the exposed rebar was subsequently repaired in October of 2010.

The second IWL examination performed after the November 2006, ILRT was completed in August 2011, in accordance with the requirements of the 2004 Edition of ASME Section XI in the second 10-year interval. The conditions observed in the 2011 inspection are only associated with Group 7 elements, which are those elements located above elevation 710 ft.-6 in. (springline) that form the shape of the dome. The maximum depth of the spalling and popout identified in 2011 was only 1 inch. The conditions identified in the 2011 inspection are still bounded by the evaluation to AEP-NRC-2020-72 Page 11 performed for the conditions identified in the 2001 and 2006 inspections, which included evaluating Group 7 elements for a loss of concrete cover up to 3 inches.

The requirements of the second 10-year interval of the CISI Program have been met for CNP Unit 2.

The most recent IWL examinations for Unit 2 were completed in July 2017, in accordance with the requirements of the 2004 Edition of ASME Section XI in the second 10-year interval. When compared with the previous IWL inspections, the new conditions observed in the 2017 inspection are only associated with Group 7 elements (spalling and popout of maximum 1 in. depth). The examination of the Unit 2 containment structure did not reveal any significant observations that could potentially affect the structural integrity or the calculated design safety margins. The subject conditions are within the bounds of the conditions that were previously identified and evaluated in the 2001, 2006, and 20_11_ CISI Program in$pections,_and the condition_ofJhe_containment concrete is being tracked under-the CISI program. The conditions that were observed in the previous IWL inspections have either been repaired or determined to be structurally acceptable.

An additional IWL examination or Appendix J inspection will be completed prior to the requested ILRT performance date.

4.5 Deficiencies Identified Consistent with the guidance provided in NEI 94-01, Revision 3-A (Reference 10), 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 a one-time extension of the CNP Unit 2 ILRT surveillance interval by eighteen months from the currently approved value of 15 years to 16.5 years. The evaluation is included as Enclosure 3 to this letter. The plant-specific risk assessment followed the guidance in NEI 94-01, Revision 3-A (Reference 10), 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 (Reference 14), the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in RG 1.200 (Reference 15) as applied to ILRT interval extensions, risk insights in support of a request for a plant's licensing basis as outlined in RG 1.174 (Reference 16), 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 (Reference 17), and the methodology used in EPRI TR 1018243, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals" (Reference 18).

4.6.2 Summary of Plant-Specific Risk Assessment Results The risk assessment associated with a one-time extension to the CNP Unit 2 ILRT surveillance interval from 15 years to 16.5 years is considered to be small since it represents a small change to the CNP Unit 2 risk profile. Details of the CNP Unit 2 risk assessment are contained in Enclosure 3

Enclosure 2 to AEP-NRC-2020-72 Page 12 to this letter. The plant-specific results for a one-time extension of the CNP Unit 2 ILR1 surveillance interval from the current 15 years to 16.5 years are summarized below.

• RG 1.174 (Reference 16) provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines "very small" changes in risk as resulting in increases in Core Damage Frequency (CDF) less than 1.0E-06/year and increases in Large Early Release Frequency (LERF) less than 1.0E-07/year. Since the ILRT does not impact CDF, the relevant criterion is LERF. The one-time increase in LERF resulting from a change in the Type A ILRT test interval from 1 in 15 years to 1 in 16.5 years is estimated as 2.96E-8/year for Unit 2 using the EPRI guidance; this value increases negligibly if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included. Therefore, the estimated change in LERF is determined to be "very small" using the acceptance guidelines of RG 1.174.

• When external event risk is included, the one-time increase in LERF resulting from a change in the Type A ILRT test interval from 1 in 15 years to 1 in 16.5 years is estimated as 1.25E-7/year for Unit 2 using the EPRI guidance. Appendix B of Enclosure 3 demonstrates that sufficient conservatisms exist in the CNP PRA models to conclude a realistic estimate of total site LERF would remain less than 1E-5/year. As such, the estimated change in LERF is determined to be "small" using the acceptance of RG 1.174.

• The effect resulting from temporarily changing the Type A test frequency to 1 in 16.5 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.0040 person-rem/year for Unit 2. NEI 94-01, Revision 3-A (Reference 10), states that a "small" population dose is defined as an increase of s 1.0 person-rem/year, ors 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. The results of this calculation meet these criteria.

Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible.

• *The one-time increase in the conditional containment failure probability (CCFP) from the 1 in 15 year interval to a 1 in 16.5 year interval is 0.11 % for Unit 2. NEI 94-01, Revision 3-A, states that increases in CCFP of s 1.5% is "small." Therefore, this increase is determined to be "small."

5.0 REGULATORY ASSESSMENT 5.1 Applicable Regulatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met.

NUREG-1831, Safety Evaluation Report Related to the License Renewal of the Donald C. Cook Nuclear Plant, Units 1 and 2, documents the NRC's technical review of l&M's license renewal application for CNP Unit 1 and Unit 2, including Aging Management Programs (AMPs) in place at CNP Unit 2. Section 3.5.2.3.1 of NUREG-1831 lists the AMPs intended to manage the aging effects of containment, and the containment components. Of those listed, three AMPs are relevant to

Enclosure 2 to AEP-NRC-2020-72 Page 13 managing containment leakage- the Containment Leakage Rate Testing Program, lnservice Inspection -ASME Section XI, Subsection IWE Program, and lnservice Inspection -ASME Section XI, Subsection IWL Program. The requirements of the Containment Leakage Rate Testing Program, which includes Type A testing, continue to be met, as the Type A test interval continues to meet the guidelines contained in NEI 94-01. The lnservice Inspection - ASME Section XI, Subsection IWE Program, which is discussed in Section 4.4.1 of this enclosure, is unaffected by this proposed amendment request. The lnservice Inspection - ASME Section XI, Subsection IWL Program, which is discussed in Section 4.4.2 of this enclosure, is unaffected by this proposed amendment request.

10 CFR 50.36( c)(3), "Surveillance requirements," states, in part, that TS shall include the "requirements relating to test, calibration or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." This proposed change revises TS 5.5.14 to add the date-related information for the next Type A test performance. Therefore, this 10 CFR 50.36 requirement continues to be met by this change.

10 CFR 50.36( c)(5), "Administrative controls," requires that "provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner," will be included in the TS. 10 CFR 50, Appendix J, Option B,Section V.B, "Implementation," requires that the implementation document used to develop a performance-based leakage testing program be included by general reference in the TS. The Appendix J Testing Program is included in the Administrative Controls section of the CNP Unit 2 TS, as TS 5.5.14, Containment Leakage Rate Testing Program. This proposed change does not remove this administrative control requirement, but simply revises TS 5.5.14, to extend the interval for performing the next Type A Test by approximately eighteen months. Therefore, this 10 CFR 50.36 requirement continues to be met by this proposed change.

10 CFR 50.54( o) requires primary reactor containments for water-cooled power reactors to be subject to the requirements of 10 CFR Part 50, Appendix J. 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.

The adoption of the Option B performance-based containment leakage rate testing for Type A, Type B, and Type C 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 Part 50, Appendix J, the test frequency is based upon an evaluation that reviewed "as-found" leakage history to determine the frequency for leakage testing which provides assurance that leakage limits will be maintained.

The one-time extension of the frequency of the containment Type A test will not affect the design, fabrication, or construction of the containment structure, and the design will continue to account for the effects of natural phenomena. The containment Type A test will continue to be done in accordance with 10 CFR 50 Appendix J using 10 CFR 50 Appendix B quality standards. The

-

• frequency of the containment Type A test is being changed in accordance with standards reviewed

Enclosure 2 to AEP-NRC-2020-72 Page 14 and approved as compliant with Appendix J. Therefore, there will be no instances where the applicable regulatory criteria are not met.

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 Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP), is submitting a request for an amendment to the Technical Specifications (TS) for CNP Unit 2. The proposed license amendment request (LAR) would allow for a one-time extension to the 15-year frequency of the CNP Unit 2 containment leakage rate test (i.e., Integrated Leak Rate Test (ILRT) or Type A test). This test is required by TS 5.5.14, Containment Leakage Rate Testing Program. The proposed one-time change would permit the current Type A Test interval of 15 years to be extended by approximately eighteen months to no later than the plant startup after the fall 2022 refueling outage.

l&M 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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment involves changes to the CNP Unit 2 containment leakage rate testing program. The proposed amendment does not involve a physical change to the plant or a change in the manner in which the plant is 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 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 modifies TS 5.5.14, Containment Leakage Rate Testing Program, to allow for a one-time extension to the containment Type A test interval. The potential consequences of extending the containment Type A test interval one time by approximately eighteen months have been evaluated by analyzing the resulting changes in risk. The increase in risk in terms of person-rem per year within 50 miles resulting from design basis accidents was estimated to be acceptably small to AEP-NRC-2020-72 Page 15 and determined to be within the guidelines published in the Nuclear Regulatory Commission (NRC)

Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 94-01, Revision 3-A.

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.

l&M has determined that the increase in conditional containment failure probability due to the proposed change would be very small. Therefore, it is concluded that the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed amendment modifies TS 5.5.14, Containment Leakage Rate Testing Program, to allow for a one-time extension to the containment Type A test interval. Containment, and the testing requirements to periodically demonstrate the integrity of containment, exist to ensure the plant's ability to mitigate the consequences of an accident and 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 amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed amendment modifies TS 5.5.14, Containment Leakage Rate Testing Program, to allow for a one-time extension to the containment Type A test interval. This amendment does not alter the manner in which safety limits, limiting safety system setpoints, or lirniting conditions for operation are determined. The specific requirements and conditions of the containment leakage rate testing program, as defined in the TS, ensure that the degree of primary containment structural integrity and leak-tightness that is considered in the plant's safety analysis is maintained. The overall containment leakage rate limit specified by the TS is maintained, and the Type A, Type B, and Type C containment leakage tests would be performed at the frequencies established in accordance with the NRG-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 containment would not degrade in a manner that is not detectable by a Type A Test. A risk assessment using the current CNP Unit 2 Probabilistic Risk Analysis model concluded that extending the Type A test interval one-time by approximately eighteen months results in a small change to the risk profile. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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. l&M concludes that the proposed amendment presents no significant hazards

Enclosure 2 to AEP-NRC-2020-72 Page 16 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 A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, 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.

6.0 CONCLUSION

In RIS 2008-27 (Reference 7), the NRC has established its position on the extension of the containment Type A test interval beyond 15 years under 10 CFR Part 50, Appendix J, Option B. The NRC will consider such extensions only under compelling circumstances. The licensee should demonstrate a sound technical justification and/or undue hardship or unusual difficulty, that the requested amendment poses minimal safety risk, acceptable plant-specific containment performance, including a plant-specific risk-informed analysis, and that the containment does not have a history of significant degradation issues.

Based on the results of previous Type A, Type B, and Type C tests performed at CNP Unit 2, along with recent IWE and IWL examinations, CNP Unit 2 containment does not have a history of significant degradation issues. The plant-specific risk analysis demonstrates that the increased risk due to an extension of approximately eighteen months to the containment Type A test is minimal. The continuation of the COVID-19 public health emergency into the spring of 2021 represents an unforeseen emergent condition, and the benefits of reducing the scope of the spring 2021 outage to mitigate the risk of transmittal and spread of COVID-19 provide a compelling basis to grant an extension.

7.0 PRECEDENT The proposed amendment incorporates into the CNP Unit 2 TS a change that is similar (i.e., an ILRT interval greater than 15 years), to the following license amendments previously approved by the NRC to extend the Type A test frequency:

• September 3, 2020 (ML20213C704), for Cook Nuclear Plant, Unit 1

• December 20, 2018 (ML18337A422), for Indian Point Nuclear Generating Station, Unit 3

• June 29, 2007 (ML071800319), for Three Mile Island Nuclear Station, Unit 1

• March 24, 2006 (ML060520032), for Seabrook Station, Unit 1

• February 9, 2006 (ML060410310), for River Bend Station, Unit 1

• December 23, 2005 (ML053190343), for St. Lucie Plant, Unit 2 to AEP-NRC-2020-72 Page 17 The proposed amendment is also similar in nature to the amendment approved on April 15, 2020 (ML20101G054 ), for Grand Gulf Nuclear Station, Unit 1, which provided a one-cycle extension to the Type A test frequency due to the need to minimize exposure of essential and non-essential personnel to the COVID-19 virus. The Grand Gulf extension did not put the test interval beyond 15 years, but does represent a case where delaying the Type A test was justified due to concerns relating to the COVID-19 public health emergency.

8.0 REFERENCES

1. Regulatory Guide 1.163, "Performance-Based Containment Leak Test Program," Revision 0, dated September 1995.

2. 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.

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

4. NEI document NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 2, dated August 31, 2007 (ADAMS Accession No. ML072970206).

5. 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 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 (ADAMS Accession No. ML081140105).

6. NEI document NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 2-A, dated November 19, 2008 (ADAMS Accession No. ML100620847).

7. 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 (ADAMS Accession No. ML080020394).

8. NEI document NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 3, dated June 9, 2011 (ADAMS Accession No. ML112920567).

9. 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 (ADAMS Accession No. ML121030286).

10. NEI document NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 3-A, dated July 31, 2012 (ADAMS Accession No. ML12221A202).

11. Letter from A. W. Dietrich, NRC, to L. J. Weber, Indiana Michigan Power Company, Donald C. Cook Nuclear Plant, Units 1 and 2 - Issuance of Amendments Re: Containment Leakage to AEP-NRC-2020-72 Page 18 Rate Testing Program (TAC Nos. MF3568 and MF3569), dated March 30, 2015 (ADAMS Accession No. ML15072A264).

12. Letter from A. W. Dietrich, NRC, to J. P. Gebbie, Indiana Michigan Power Company, "Donald C. Cook Nuclear Plant, Units 1 and 2 - Issuance of Amendments Re: Adoption of TSTF-490, Rev. 0, "Deletion of E-Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification," and Implementation of Full-Scope Alternative Source Tenn (CAC Nos. MF5184 and MF5185)," dated October 20, 2016 (ADAMS Accession No. ML16242A111 ). *

13. NUREG-1493, "Perfonnance-Based Containment Leak-Test Program," dated September 1995.

14. Interim Guidance for Perfonning Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals, Revision 4, developed for NEI by EPRI and Data Systems and Solutions, dated November 2001.

15. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, dated March 2009.

16. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 3, dated January 2018.

17. Letter from C. H. Cruse, Calvert Cliffs Nuclear Power Plant, to NRC Document Control Desk, "Response to Request for Additional lnfonnation Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension," dated March 27, 2002 (ADAMS Accession No. ML020920100).

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

Enclosure 3 to AEP-NRC-2020-72 Evaluation of Risk Significance of Short-Term ILRT Extension

0 JENSEN HUGHES Advancing the Science of Safety Donald C. Cook Nuclear Plant:

Evaluation of Risk of Short-Term ILRT Extension - Unit 2 PRA-QNT-011 Prepared for:

~ INDIANA

__,MICHIGAN 0

POWER Aumt ofAm0rfcan El8ctrlc Power Donald C. Cook Nuclear Plant Indiana Michigan Power Company Project Number: 1EAT1V046 Project

Title:

Short-Term ILRT Extension Revision: 0 Name and Date D1g1taJly signed by Justin Sattler Preparer Justin Sattler Date: 2020.12.04 08:37:11-06'00' Reviewer Matthew Johnson Review Method Design Review ~ Alternate Calculation

~-byEnolhomobwy I r-DN C=fUS, E ~ ccm, c:>-Jll"l:Ntl Hur,.tl.N, Approver. Eric Thornsbury Enc Thornsbu ~

I -=S~oo~~

Revision 0 Page 2 of 42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension REVISION RECORD

SUMMARY

Revision Revision Summary 0 lnrbal Issue Revision 0 Page 3 of42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension TABLE OF CONTENTS 1.0 PURPOSE ...................................................................................................................... 5 2.0 SCOPE ........................................................................................................................... 5

3.0 REFERENCES

............................................................................................................... 7 4.0 ASSUMPTIONS AND LIMITATIONS ............................................................................. 10 5.0 METHODOLOGY AND INPUTS .................................................................................... 11 5.1 General Resources Available ...................................................................................... 11 5.2 Plant Specific Inputs ................................................................................................... 12 5.3 Impact of Extension on Detection of Component Failures that Lead to Leakage (Small and Large) ...............................................................................................................14 6.0 CALCULATIONS ...........................................................................................................15 6.1 Step 1 - Quantify the Risk in Terms of Frequency per Reactor Year .......................... 16 6.2 Step 2 - Develop Plant-Specific Person-Rem Dose (Population Dose) ....................... 19 6.3 Step 3 - Evaluate Risk Impact of Temporarily Extending Type A Test Interval from 15 to 16.5 Years ...........................................................................................................20 6.4 Step 4 - Determine the Change in Risk in Terms of LERF .......................................... 22 6.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability .... 23 6.6 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage ...... 24 6.7 Impact from External Events Contribution ...................................................................27 7.0 RESULTS ......................................................................................................................29

8.0 CONCLUSION

S AND RECOMMENDATIONS .............................................................. 30 A. PRA Technical Adequacy for ILRT ................................................................................32

8. Total Site Risk Discussion .............................................................................................33 Revision 0 Page 4 of42

PRA-QNT-011 Evaluation of Risk Slgntflcance of Short-Term ILRT Extension 1.0 PURPOSE The purpose of this analysis is to provide a risk assessment of temporarily extending the currently allowed containment Type A Integrated Leak Rate Test (ILRT) interval by 18 months from 15 years to 16.5 years for Unit 2 of the Donald C. Cook Nuclear Plant (CNP). The risk assessment follows the guidelines from Nuclear Energy Institute (NEI) 94-01, Revision 3-A

[Reference 1], 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 [Reference 3], the Nuclear Regulatory Commission (NRC) regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in Regulatory Guide (RG) 1.200 as applied to ILRT interval extensions, risk insights in support of a request for a plant's licensing basis as outlined in RG 1.174 [Reference 4], 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 [Reference 5], and the methodology used in Electric Power Research Institute (EPRI) 1018243, which is Revision 2-A of EPRI 1009325

[Reference 24].

• 2.0 SCOPE Option B to Appendix J. Section 11.0 of NEI 94-01 states that NUREG-1493, "Performance-Based Containment Leak Test Program," September 1995 [Reference 6], provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement the NRC's rulemaking basis, NEI undertook a similar study. The results of that study are documented in EPRI Research Project TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals" [Reference 2].

The NRC report on performance-based leak testing, NUREG-1493, analyzed the effects of containment leakage on the health and safety of the public and the benefits realized from the containment leak rate testing. In that analysis, it was determined that for a representative PWR plant (i.e., Surry), containment isolation failures contribute less than 0.1 % to the latent risks from reactor accidents. Consequently, it is desirable to show that extending the ILRT interval will not

-lead to a substantial increase in risk from containment isolation failures for CNP.

NEI 94-01 Revision 3-A supports using EPRI Report No. 1018243, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," for performing risk impact assessments in support of ILRT extensions [Reference 24]. The Guidance provided in Appendix H of EPRI Report No. 1018243 builds on the EPRI Risk Assessment methodology, EPRI TR-104285. This methodology is followed to determine the appropriate risk information for use in evaluating the impact of the proposed ILRT changes.

It should be noted that containment leak-tight integrity is also verified through periodic in-service inspections conducted in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI. More specifically, Subsection IWE provides the rules and requirements for in-service inspection of Class MC pressure-retaining components and their integral attachments, and of metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments in light-water cooled plants. Furthermore, NRC regulations 10 CFR 50.55a(b)(2)(ix)(E) require licensees to conduct visual inspections of the accessible areas of the interior of the containment. The associated change to NEI 94-01 will require that visual examinations be conducted during at least three other outages, and in the outage during which the ILRT is being

_conducted. These requirements will not be changed as a result of the extended ILRT interval. In Revision 0 Page 5 of42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension addition, Appendix J, Type B local leak tests performed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, and gaskets are also not affected by the change to the Type A test frequency.

The acceptance guidelines in RG 1.174 are used to assess the acceptability of this temporary extension of the Type A test interval beyond that established during the Option B rulemaking of Appendix J. RG 1.174 defines "very small" changes in the risk-acceptance guidelines as increases in Core Damage Frequency (CDF) less than 1o-8 per reactor year and increases in Large Early Release Frequency (LERF) less than 10-7 per reactor year. Since containment accident pressure is not credited in support of ECCS performance, the Type A test does not impact CDF; therefore, the relevant criterion is the change in LERF. RG 1.174 also defines "small" changes in LERF as below 1o-8 per reactor year. RG 1.174 discusses defense-in-depth and encourages the use of risk analysis techniques to help ensure and show that key principles, such as the defense-in-depth philosophy, are met Therefore, the increase in the Conditional Containment Failure Probability (CCFP), which helps ensure the defense-in-depth philosophy is maintained, is also calculated.

Regarding CCFP, changes of up to 1.1 % have been accepted by the NRC for the one-time requests for extension of ILRT intervals. In context, it is noted that a CCFP of 1/10 (10%) has been approved for application to evolutionary light water designs. Given these perspectives, a change in the CCFP of up to 1.5% is assumed to be small.

In addition, the total annual risk (person-rem/year population dose) is examined to demonstrate the relative change in this parameter. While no acceptance guidelines for these additional figures of merit are published, examinations of NUREG-1493 and Safety Evaluations (SEs) for one-time interval extension (summarized in Appendix G of Reference 24) indicate a range of incremental increases in population dose that have been accepted by the NRC. The range of incremental population dose increases is from S0.01 to 0.2 person-rem/year and/or 0.002% to 0.46% of the total accident dose. The total doses for the spectrum of all accidents (NUREG-1493 [Reference 6], Figure 7-2) result in health effects that are at least two orders of magnitude less than the NRC Safety Goal Risk. Given these perspectives, a small population dose is defined as an increase from the baseline interval dose of :::.1.0 person-rem per year or 1% of the total dose, whichever is less restrictive for the risk impact assessment of the proposed extended ILRT interval [Reference 1].

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PRA-QNT-011 Evaluation of Risk Significance of Short-Term ILRT Extension

3.0 REFERENCES

The following references were used in this calculation:

1. Revision 3-A to Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, NEI 94-01, July 2012.

2. Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, EPRI, Palo Alto, CA EPRI TR-104285, August 1994.

3. Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals, Revision 4, developed for NEI by EPRI and Data Systems and Solutions, November 2001.

4. An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 3, January 2018.

5. Response to Request for Additional Information Concerning the Ucense Amendment Request for a One-Time Integrated Leakage Rate Test Extension, Letter from Mr. C. H.

Cruse (Calvert Cliffs Nuclear Power Plant) to NRC Document Control Desk, Docket No.

50-317, March 27, 2002.

6. Performance-Based Containment Leak-Test Program, NUREG-1493, September 1995.

7. Evaluation of Severe Accident Risks: Surry Unit 1, Main Report NUREG/CR-4551, SAND86-1309, Volume 3, Revision 1, Part 1, October 1990.

8. Letter from R. J. Barrett (Entergy) to U.S. Nuclear Regulatory Commission, IPN-01-007, Ja11uary 18, 2001.

9. United States Nuclear Regulatory Commission, Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: Frequency of Performance-Based Leakage Rate Testing (TAC No. MB0178), April 17, 2001.

10. Impact of Containment Building Leakage on LWR Accident Risk, Oak Ridge National Laboratory, NUREG/CR-3539, ORNL/TM-8964, April 1984.

11. Reliability Analysis of Containment Isolation Systems, Pacific Northwest Laboratory, NUREG/CR-4220, PNL-5432, June 1985.

12. Technical Findings and Regulatory Analysis for Generic Safety Issue 11.E.4.3

'Containment Integrity Check', NUREG-1273, April 1988.

13. Review of Ught Water Reactor Regulatory Requirements, Pacific Northwest Laboratory, NUREG/CR-4330, PNL-5809, Volume 2, June 1986.

14. Shutdown Risk Impact Assessment for Extended Containment Leakage Testing Intervals Utilizing ORAM', EPRI, Palo Alto, CA, TR-105189, Final Report, May 1995.

15. Severe Accident Risks: An Assessment for Ftve U. S. Nuclear Power Plants, NUREG-1150, December 1990.

16. United States Nuclear Regulatory Commission, Reactor Safety Study, WASH-1400, October 1975.

17. Calculation PRA-NB-QU, "Internal Events Quantification Notebook.a Revision 6, June 2020.

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PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension

18. Calculation PRA-NB-LER, "Large Early Release Frequency Notebook," Revision 3, June 2020.

19. Calculation PRA-ILRT, "Risk Impact Assessment For Permanently Extending Containment Type A Test Interval," Revision 2, March 2014.

20. Anthony R. Pietrangelo, One-time extensions of containment integrated leak rate test interval - additional information, NEI letter to Administrative Points of Contact, November 30, 2001.

21. Letter from J. A. Hutton (Exelon, Peach Bottom) to U. S. Nuclear Regulatory Commission, Docket No. 50-278, License No. DPR-56, LAR-01-00430, dated May 30, 2001.

22. Risk Assessment for Joseph M. Farley Nuclear Plant Regarding ILRT (Type A)

Extension Request, prepared for Southern Nuclear Operating Co. by ERIN Engineering and Research, P0293010002-1929-030602, March 2002.

23. Letter from D. E. Young (Florida Power, Crystal River) to U.S. Nuclear Regulatory Commission, 3F0401-11, dated April 25, 2001.

24. Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325, EPRI, Palo Alto, CA, 1018243, October 2008.

25. Risk Assessment for Vogtfe Electric Generating Plant Regarding the ILRT (Type A)

Extension Request, prepared for Southern Nuclear Operating Co. by ERIN Engineering and Research, February 2003.

26. Perspectives Gained from the IPEEE Program, USNRC, NUREG-1742, April 2002.

27. Report PRA-NB-SPRA-QU, "SPRA Model Quantification Notebook," Revision 2, October 2019.

28. Calculation PRA-NB-FIRE-FQ, "Fire PRA Model Quantification Notebook," Revision 3, October 2019.

29. NTS 2003 009 REP, Revision O," Cook Nuclear Plant Severe Accident Mitigation Alternatives Analysis," October 20, 2003.

30. 1-EHP-4030-134-202, "Integrated Leak Rate Test," Revision 3.

31. NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition.

32. "Individual Plant Examination of External Events Summary Report," April 1992.

33. NUREG/CR-6850 Supplement 1, "Fire Probabilistic Risk Assessment Methods Enhancements," September 2010.

34. Technical Letter Report ML112070867, Containment Liner Corrosion Operating Experience Summary, Revision 1, August 2011.

35. NUREG-0800, Section 6.5.2, "Containment Spray as a Fission Product Cleanup System," Revision 3.

36. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, March 2009.

37. ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," 2009.

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PRA-QNT-011 Evaluation of Risk Significance of Short-Term ILRT Extension

38. NEI 05-04, "Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard," Revision 2, November 2008.

39. ACUBE 2.0 Software Manual, EPRI Report 3002003169, December 2014.

40. NEI Letter to USNRC, "Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os)," February 21, 2017, Accession Number ML17086A431.

41. USN RC Letter to Mr. Greg Krueger (NEI), "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 7-12, and 12-13, Close Out of Facts and Observations (F&Os)," May 3, 2017, Accession Number ML17079A427.

42. NUREG-2169, "Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database: United States Fire Event Experience Through 2009," January 2015.

43. LTR-RAM-11-10-041, "Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for the DC Cook Fire Probabilistic Risk Assessment," July 20, 2010.

44. Erin Engineering Report #D0403140002, "D.C. Cook Focused Scope Peer Review for Fire PRA," November 2015.

45. PWROG-17027-P, "Focused Scope Peer Review of the DC Cook Internal Fire Probabilistic Risk Assessment," July 2017.

46. PWROG-18062-P, "Peer Review of the D.C. Cook 1&2 Seismic Probabilistic Risk Assessment," Revision 0, January 2019.

47. Report AEPDCC-00058-REPT-001, "SPRA F&O Independent Assessment and Focused-Scope Peer Review, Donald C. Cook Nuclear Plant Units 1 and 2," Revision A.

48. AEP-NRC-2019-56, "Seismic Probabilistic Risk Assessment in Response to Near Term Task Force Recommendation 2.1: Seismic," November 4, 2019.

49. PWROG-14003-NP, "Implementation of FLEX Equipment in Plant-Specific PRA Models," Revision 2, August 2016.

50. PWROG-18043-P, "FLEX Equipment Data Collection and Analysis," Revision 0, February 2020.

51. PWROG-16034-P, "Implementation Guidance for Risk Aggregation Pilot," Revision 0, July 2017.

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PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension 4.0 ASSUMPTIONS AND LIMITATIONS The following assumptions were used in the calculation:

The acceptability (i.e., technical adequacy) of the CNP PRA [Reference 17] is either consistent with the requirements of Regulatory Guide 1.200, or where gaps exist, the gaps have been addressed, as detailed in Appendix A The CNP Level 1 and 2 internal events PRA models provide representative results.

It is appropriate to use the CNP internal events PRA model to effectively describe the risk change attributable to the ILRT extension. An analysis is performed in Section 6.7 to show the effect of including external event models for the ILRT extension. The Seismic PRA [Reference 28] and Fire PRA [Reference 18] are used for this analysis.

Accident classes describing radionuclide release end states are defined consistent with EPRI methodology [Reference 24].

The representative containment leakage for Class 1 sequences is 1La. Class 3 sequences account for increased leakage due to Type A inspection failures.

The representative containment leakage for Class 3a sequences is 1Ola based on the previously approved methodology performed for Indian Point Unit 3 [Reference 8, Reference 9].

The representative containment leakage for Class 3b sequences is 1OOLa based on the guidance provided in EPRI Report No. 1018243 [Reference 24].

The sequences of Class 3b can be very conservatively categorized as LERF based on the previously approved methodology [Reference 8, Reference 9].

The impact on population doses from containment bypass scenarios is not altered by the proposed ILRT extension, but is accounted for in the EPRI methodology as a separate entry for comparison purposes. Since the containment bypass contribution to population dose is fixed, no changes in the conclusions from this analysis will result from this separate categorization.

The reduction in ILRT frequency does not impact the reliability of containment isolation valves to close in response to a containment isolation signal [Reference 24].

While precise numbers are maintained throughout the calculations, some values have been rounded when presented in this report. Therefore, rounding differences may result in differences in table summations.

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PRA-QNT-011 Evaluatlon of RJsk Significance of Short-Tenn ILRT Extension 5.0 METHODOLOGY AND INPUTS This section summarizes the general resources available as input (Section 5.1) and the plant specific resources required (Section 5.2).

5.1 General Resources Available Various industry studies on containment leakage risk assessment are briefly summarized here:

1. NUREG/CR-3539 [Reference 1OJ

2. NUREG/CR-4220 [Reference 11]

3. NUREG-1273 [Reference 12]

4. NUREG/CR-4330 [Reference 13]

5. EPRI TR-105189 [Reference 14]

6. NUREG-1493 [Reference 6]

7. EPRI TR-104285 [Reference 2]

8. NUREG-1150 [Reference 15] and NUREG/CR-4551 [Reference 7]

9. NEI Interim Guidance [Reference 3, Reference 20]

10. Calvert Cliffs liner corrosion analysis [Reference 5]

11. EPRI Report No. 1009325, Revision 2-A (EPRI 1018243), Appendix H [Reference 24]

This first study is applicable because it provides one basis for the threshold that could be used in the Level 2 PRA for the size of containment leakage that is considered significant and is to be included in the model. The second study is applicable because it provides a basis of the probability for significant pre-existing containment leakage at the time of a core damage accident. The third study is applicable because it is a subsequent study to NUREG/CR-4220 that undertook a more extensive evaluation of the same database. The fourth study provides an assessment of the impact of different containment leakage rates on plant risk. The.fifth study provides an assessment of the impact on shutdown risk from ILRT test interval extension. The sixth study is the NRC's cost-benefit analysis of various alternative approaches regarding extending the test intervals and increasing the allowable leakage rates for containment integrated and local leak rate tests. The seventh study is an EPRI study of the impact of extending ILRT and local leak rate test (LLRT) intervals on at-power public risk. The eighth study provides an ex-plant consequence analysis for a 50-mile radius surrounding a plant that is used as the basis for the consequence analysis of the ILRT interval extension for CNP. The ninth study includes the NEI recommended methodology (promulgated in two letters) for evaluating the risk associated with obtaining a one-time extension of the ILRT interval. The tenth study addresses the impact of age-related degradation of the containment liners on ILRT evaluations. Finally, the eleventh study builds on the previous work and includes a recommended methodology and template for evaluating the risk associated with a permanent 15-year extension of the ILRT interval.

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PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension 5.2 Plant Specific Inputs The plant-specific information used to perform the CNP ILRT Extension Risk Assessment includes the following:

CDF and LERF Model results [Reference 17, Reference 18]

Dose within a 50-mile radius [Reference 19, Reference 29]

CNP Model The Internal Events PRA Model that is used for CNP is characteristic of the as-built plant. The current Level 1, LERF, and Level 2 model is a linked fault tree model [Reference 17]. The CDF is 2.73E-5/year for Unit 2; the LERF is 1.78E-6/year for Unit 2 [Reference 17]. Table 5-1 and Table 5-2 provide a summary of the Internal Events CDF and LERF results for the CNP PRA Model.

The total Fire CDF is 6.81 E-5/year for Unit 2; the total Fire LERF is 3.70E-6/year for Unit 2

[Reference 18]. The total Seismic CDF is 2.38E-5/year for Unit 2; the total Seismic LERF is 5.36E-6/year [Reference 28]. Refer to Section 6. 7 for further details on external events as they pertain to this analysis.

Table 5 Internal Events CDF Internal Events Unit 2 Frequency (per year)

Internal Floods 7.07E-06 Transients 1 58E-05 Main Steam Break Inside Containment 1.19E-07 LO~As 3.25E-06 ISLOCA 3.23E-07 SGTR 2.29E-07 RPV Rupture 2.65E-08 Loss of Offsite Power (LOOP) 4.70E-07 Total Internal Events CDF 2.73E-05 Table 5 Internal Events LERF Internal Events Unit 2 Frequency (per year)

Internal Floods 3 45E-07 Transients 5.12E-07 Main Steam Break Inside Containment 2.68E-09 LOCAs 6.63E-07 ISLOCA 1 47E-07 SGTR 5 33E-08 RPV Rupture 2.B0E-08 LOOP 3.61E-08 Total Internal Events LERF 1.78E-06 Revision 0 Page 12 of 42

PRA-QNT-011 Evaluation of Risk Slgntflcance of Short-Term ILRT Extension Population Dose Calculations The population dose calculation was reported m the permanent 15-year ILRT extension

[Reference 19], which used data from the SAMA [Reference 29]. Table 5-3 presents these dose exposures.

Table 5 Population Dose EPRI Category Dose (person-rem)

Class 1 1.01E+03 Class 2 3 84E+06 Class 7 3.84E+06 Class 8 9 68E+06 Release Category Definitions Table 5-4 defines the accident classes used in the ILRT extension evaluation, which is consistent with the EPRI methodology [Reference 24]. These containment failure classifications are used m this analysis to determine the risk impact of extending the Containment Type A test interval, as de$cribed in Section 6.0 of this report.

Table 5 EPRI Containment Failure Classification [Reference 24]

Class Description Containment remains intact including accident sequences that do not lead to containment failure in the long term. The release of fission products (and attendant consequences) IS determined by the maximum allowable leakage rate values La, under Appendix J for that plant.

Containment 1solabon failures (as reported in the Individual Plant Exam1nat1ons) including those accidents 2

,n which there Is a failure to isolate the containment.

Independent (or random) isolation failures include those accidents in which the pre-eXJstmg isolation 3

failure to seal (1 e , provide a leak-bght containment) IS not dependent on the sequence in progress Independent (or random) isolation failures include those accidents in which the pr~XJsting isolation failure to seal Is not dependent on the sequence in progress. This class Is s1m1lar to Class 3 1solabon 4

failures, but Is applicable to sequences involving Type B tests and their potenbal failures These are the Type B-tested components that have isolated, but exh1brt excessive leakage.

Independent (or random) isolation failures including those acadents in which the pre-exIstIng isolation 5 failure to seal IS not dependent on the sequence in progress This class Is s1m1lar to Class 4 Isolabon failures, but Is applicable to sequences Involvmg Type C test and their potenbal failures Containment isolation failures including those leak paths covered in the plant test and maintenance 6

requirements or venfied per In-se1"V1ce inspection and testing (ISI/ISD program.

Accidents involving containment failure Induced by severe accident phenomena. Changes in AppendlX J 7

testing requirements do not impact these acadents.

Accidents In which the containment Is bypassed (either as an inrbal condrbon or induced by phenomena) 8 are included in Class 8 Changes In Appendix J testing requirements do not impact these aCC1dents.

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PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension 5.3 Impact of Extension on Detection of Component Failures that Lead to Leakage (Small and Large)

The ILRT can detect a number of component failures such as liner breach, failure of certain bellows arrangements, and failure of some sealing surfaces, which can lead to leakage. The proposed ILRT test interval extension may influence the conditional probability of detecting these types of failures. To ensure that this effect is properly addressed, the EPRI Class 3 accident class, as defined in Table 5-4, is divided into two sub-classes, Class 3a and Class 3b, representing small and large leakage failures, respectively.

The probability of the EPRI Class 3a and Class 3b failures is determined consistent with the EPRI Guidance [Reference 24]. For Class 3a, the probability is based on the maximum likelihood estimate of failure (arithmetic average) from the available data (i.e., 2 "small" failures in 217 tests leads to 2 / 217 = 0.0092). For Class 3b, the probability is based on the Jeffreys non-infonnatIve prior for no "large" failures in 217 tests (i.e., 0.5 / (217+1) = 0.0023).

In a follow-up letter [Reference 20] to their ILRT guidance document [Reference 3], NEI issued additional information concerning the potential that the calculated delta LERF values for several plants may fall above the "very small change" guidelines of the NRG Regulatory Guide 1.174

[Reference 4]. This additional NEI information includes a discussion of conservatisms in the quantitative guidance for b.LERF. NEI describes ways to demonstrate that, using plant-specific calculations, the .l\LERF is smaller than that calculated by the simplified method.

The supplemental information states:

The methodology employed for determining LERF (Class 3b frequency) involves conseNatively multiplying the GDF by the failure probability for this class (3b) of accident. This was done for simplicity and to maintain conservatism. However, some plant-specific accident classes leading to core damage are likely to include individual sequences that either may already (independently) cause a LERF or could never cause a LERF, and are thus not associated with a postulated large Type A containment leakage path (LERF). These contributors can be removed from Class 3b in the evaluation of LERF by multiplying the Class 3b probability by only that portion of GDF that may be impacted by Type A leakage.

The application of this additional guidance to the analysis for CNP, as detailed in Sect.Jon 6.0, involves subtracting LERF risk (except the risk associated with a pre-existing leak) from the GDF that is applied to Class 3b because this portion of LERF is unaffected by containment integrity. To be consistent, the same change is made to the Class 3a GDF, even though these events are not considered LERF.

Consistent with the NEI Guidance [Reference 3], the change in the leak detection probability can be estimated by comparing the average time that a leak could exist without detection. For example, the average time that a leak could go undetected with a three-year test interval is 1.5 years (3 years I 2), the average time that a leak could exist without detection for a 15-year test interval is 7.5 years (15 years/ 2), and the average time that a leak could exist without detection for a 16.5-year interval is 8.25 years (16.5 years/ 2). The change from a three-year test interval to a 15-year test interval would lead to a non-detection probability that is a factor of 5.0 (7.5/1.5) higher for the probability of a leak that is detectable only by ILRT testing. Correspondingly, an extension of the ILRT interval to 16.5 years can be estimated to lead to a factor of 5.5

((16.5/2)/1.5) increase in the non-detection probability of a leak.

It should be noted that using the methodology discussed above is very conservative compared to previous submittals (e.g., the IP3 request for a one-time ILRT extension that was approved by the NRC [Reference 91) because it does not factor in the possibility that the failures could be Revision 0 Page 14 of 42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension detected by other tests (e.g., the Type B local leak rate tests that will still occur). Eliminating this possibility conservatively over-estimates the factor increases attributable to the ILRT extension.

6.0 CALCULATIONS The application of the approach based on the guidance contained in EPRI 1018243 [Reference 24] and previous risk assessment submittals on this subject [References 5, 8, 21, 22, and 23]

have led to the following results. The results are displayed according to the eight accident classes defined in the EPRI report, as described in Table 6-1.

The analysis performed examined CNP-specific accident sequences in which the containment remains intact or the containment is impaired. Specifically, the breakdown of the severe accidents, contributing to risk, was considered in the following manner:

Core damage sequences in which the containment remains intact initially and in the long term (EPRI 1018243, Class 1 sequences [Reference 241).

Core damage sequences in which containment integrity is impaired due to random isolation failures of plant components other than those associated with Type B or Type C test components. For example, liner breach or bellow leakage (EPRI 1018243, Class 3 sequences [Reference 241).

Accident sequences involving containment bypassed (EPRI 1018243, Class 8 sequences [Reference 241), large containment isolation failures (EPRI 1018243, Class 2 sequences [Reference 241), and small containment isolation "failure-to-seal" events (EPRI 1018243, Class 4 and 5 sequences [Reference 241) are accounted for in this evaluation as part of the baseline risk profile. However, they are not affected by the ILRT frequency change.

Class 4 and 5 sequences are impacted by changes in Type B and C test intervals; therefore, changes in the Type A test interval do not impact these sequences.

Table 6 EPRI Accident Class Definitions Accident Classes (Containment Release Type) Description No Containment Failure 2 Large lsolabon Failures (Failure to Close) 3a Smail Isolation Failures (Liner Breach) 3b Large Isolation Failures (Liner Breach) 4 Smail lsolabon Failures (Failure to Seal - Type 8) 5 Smail Isolation Failures (Failure to Seal - Type C) 6 Other Isolation Failures (e.g., Dependent Failures) 7 Failures Induced by Phenomena (Early and Late) 8 Bypass (SGTR and lnterfacmg System LOCA)

CDF All CET End States (Including Very Low and No Release)

The steps taken to perform this risk assessment evaluation are as follows:

Step 1 - Quantify the baseline risk in terms of frequency per reactor year for each of the accident classes presented in Table 6-1.

Step 2 - Develop plant-specific person-rem dose (population dose) per reactor year for each of the eight accident classes.

Step 3 - Evaluate risk impact of a one-time extension of the Type A test interval from 1 in 15 Revision 0 Page 15 of 42

PRA-QNT-011 Evaluation of RJsk Significance of Short-Tenn ILRT Extension years to 1 in 16.5 years.

Step 4 - Determine the change in risk in terms of Large Early Release Frequency (LERF) in accordance with RG 1.174 [Reference 4].

Step 5 - Determine the impact on the Conditional Containment Failure Probability (CCFP).

6.1 Step 1 - Quantify the Risk in Terms of Frequency per Reactor Year As previously described, the extension of the Type A interval does not influence those accident progressions that involve large containment isolation failures, Type B or Type C testing, or containment failure induced by severe accident phenomena.

For the assessment of ILRT impacts on the risk profile, the potential for pre-existing leaks is included in the model (these events are represented by the Class 3 sequences in EPRI 1018243 [Reference 24D. The question on containment integrity was modified to include the probability of a liner breach or bellows failure (due to excessive leakage) at the time of core damage. Two failure modes were considered for the Class 3 sequences. These are Class 3a (small breach) and Class 3b (large breach).

The frequencies for the severe accident classes defined in Table 6-1 were developed for CNP by first determining the frequencies for Classes 1, 2, 6, 7, and 8.

Table 6-2 presents the grouping of each release category in EPRI Classes based on the associated description. Table 6-3 provides a summary of the accident sequence frequencies that can lead to radionuclide release to the public and have been derived consistent with the NEI Interim Guidance [Reference 3] and the definitions of accident classes and guidance provided in EPRI Report No. 1018243 [Reference 24]. Adjustments were made to the Class 3b and hence Class 1 frequencies to account for the impact of undetected corrosion of the steel liner per the methodology described in Section 6.6.

Class 3 Sequences. This group consists of all core damage accident progression bins for which a pre-existing leakage in the containment structure (e.g., containment liner) exists that can only be detected by performing a Type A ILRT. The probability of leakage detectable by a Type A ILRT is calculated to determine the impact of temporarily extending the testing interval. The Class 3 calculation is divided into two classes: Class 3a is defined as a small liner breach (La <

leakage < 1Ola), and Class 3b is defined as a large liner breach (1 Ola < leakage < 1OOLa).

Data reported in EPRI 1018243 [Reference 24] states that two events could have been detected only during the performance of an ILRT and thus impact risk due to change in ILRT frequency.

There were a total of 217 successful ILRTs during this data collection period. Therefore, the probability of leakage is determined for Class 3a as shown in the following equation:

2 Pclass3a = 217 = 0.0092 Multiplying the CDF by the probability of a Class 3a leak yields the Class 3a frequency contribution in accordance with guidance provided in Reference 24. As described in Section 5.3, additional consideration is made to not apply failure probabilities on those cases that are already LERF scenarios. Therefore, these LERF contributions from CDF are removed (except the LERF associated with a pre-existing leak). The frequency of a Class 3a failure for a 3 in 10 year ILRT interval is calculated by the following equation (ratios of intervals are subsequently applied to calculate frequencies for longer ILRT intervals):

Frequzc1ass3a = Pclass3a * ( CDF - (LERF - LERF1ea1c))

2

= 217 * (2.73E (1.78E-6-3.06E-7)) = 2.38E-7 Revision 0 Page 16 of 42

PRA-QNT-011 Evaluatlon of Risk Significance of Short-Tenn ILRT Extension The risk contribution is changed based by a factor of 15/3 and 16.5/3 (for the 1 in 15 year and 1 in 16.5 year cases, respectively) compared to the 3 in 10 year ILRT interval case values. The Class 3a frequencies are calculated as follows:

15 2 15 2 Frequ2czass

, 3a15yr = 3

• -217

- * (CDP - (LERF - LERFzeak)) = 3

• -217-

• 2.SBE-5 = 1.19E-6 16.5 2 Frequ2class3a16.5yr = - 3 * -217 * ( CDP - )

(LERF - LERFzeak) = - 3 16.5 2

• -217

• 2.SSE-5 = 1.31E-6 In the database of 217 ILRTs, there are zero containment leakage events that could result in a large early release. Therefore, the Jeffreys non-informative prior is used to estimate a failure rate and is illustrated in the following equations:

• Number of Failures+ 1/2 Jeffreys Failure Probability= b f Num er o Tests+ 1 0 + 1/2 217 + l = 0.0023 Pclass3b =

The frequency of a Class 3b failure for a 3 in 10 year ILRT interval is calculated by the following equation (ratios of intervals are subsequently applied to calculate frequencies for longer ILRT intervals):

Frequ2czass3b = Pc1ass3b * (CDP - (LERF - LERF1eak))

=~218

• (2.73E (1.78E-6-3.06E-7)) = 5.92E-8 The risk contribution is changed based by a factor of 15/3 and 16.5/3 (for the 1 in 15 year and 1 in 16.5 year cases, respectively) compared to the 3 in 10 year ILRT interval case values. The Class 3b frequencies are calculated as follows:

Frequ2czass3b15yr =

15 3

• ~ 218

• (CDF - (LERF - LERF1eak)) =

15 3

• ~218

• 2.SSE-5 = 2.96E-7 16 5

• ~

16 5 Frequ2class3b16 *5yr = 3

" * -1....

218

• (CDF - (LERF - LERFzeak)) = 3 218

• 2.SBE-5 = 3.25E-7 For this analysis, the associated containment leakage for Class 3a is 1 Ola and for Class 3b is 1OOLa. These assignments are consistent with the guidance provided in Reference 24.

Class 1 Sequences. This group consists of all core damage accident progression bins for which the containment remains intact (modeled as Technical Specification Leakage). The Intact frequency for internal events is 2.58E-5 for Unit 2 [Reference 17]. The EPR.I Accident Class 1 frequency is then adjusted by subtracting the EPRI Class 3a and 3b frequency (to preserve total CDF), calculated below.

Frequ2czass1 = Frequuntact - (Frequ2class3a - Frequ2class3b)

Class 2 Sequences. This group consists of accident progression bins with large containment isolation failures (except the LERF associated with a pre-existing leak). The large isolation failure is in internal events cutsets that contribute 8.80% of LERF for Unit 2. Multiplying by LERF, the EPRI Accident Class 2 frequency is 1.57E-7 for Unit 2, as shown in Table 6-2.

C!ass 4 Sequences. This group consists of all core damage accident progression bins for which containment isolation failure-to-seal of Type B test components occurs. Because these failures are detected by Type B tests which are unaffected by the Type A ILRT, this group is not evaluated any further in the analysis, consistent with approved methodology.

Class 5 Sequences. This group consists of all core damage accident progression bins for which Revision 0 Page 17 of 42

PRA-QNT..011 Evaluation of Risk Significance of Short-Tenn ILRT Extension a containment isolation failure-to-seal of Type C test components occurs. Because the failures are detected by Type C tests which are unaffected by the Type A ILRT, this group is not evaluated any further in this analysis, consistent with approved methodology.

Class 6 Sequences. These are sequences that involve core damage accident progression bins for which a failure-to-seal containment leakage due to failure to isolate the containment occurs.

These sequences are dominated by misalignment of containment isolation valves following a tesUmaintenance evolution. All other failure modes are bounded by the Class 2 assumptions.

This accident class is also not evaluated further.

Class 7 Sequences. This group consists of all core damage accident progression bins in which containment failure is induced by severe accident phenomena (e.g., overpres$ure). This frequency is calculated by subtracting the Class 1, 2, and 8 frequencies from the total CDF. For this analysis, the frequency is determined from the EPRI Accident Class 7 frequency listed in Table 6-2.

Class 8 Sequences. This group consists of all core damage accident progression bins in which containment is bypassed via interfacing-systems loss of coolant accident (ISLOCA) or steam generator tube rupture (SGTR). The ISLOCA initiators are in internal events cutsets that contribute 8.24% of LERF for Unit 2. The SGTR initiator is in internal events cutsets that contribute 2.99% of LERF for Unit 2. Thus, the total EPRI Accident Class 8 frequency is the summation of the ISLOCA and SGTR frequencies, 2.00E-7 for Unit 2, as shown in Table 6-2 and Table 6-3.

Table 6 Accident Class Frequencies EPRI Category Unit 2 Frequency (/yr)

Class 1 2.58E-05 Class 2 1.57E-07 Class 7 1 12E-06 Class 8 (SGTR) 5.33E-08 Class 8 (ISLOCA) 1.47E-07 Total (CDF) 2 73E-05 Table 6 Risk Proflle for 3 In 10 Year ILRT Class Description Unit 2 Frequency (/yr) 1 No containment failure 2.55E-05 2 2 Large containment isolation failures 1.57E-07 3a Small isolation failures (liner breach) 2.38E-07 3b Large isolation failures (liner breach) 5.92E-08 4 Small isolation failures - failure to seal (type 8) E1 5 Small isolation failures - failure to seal (type C) E1 6 Containment isolation failures (dependent failure, personnel errors) E1 7 Severe accident phenomena induced failure (early and late) 1.12E-06 8 Containment bypass 2 00E-07 Total 2.73E-05

1. E represents a probab1hsbcally insignificant value or a Class that Is unaffected by the Type A ILRT

2. The Class 3a and 3b frequene1es are subtracted from Class 1 to preserve total CDF.

Revision 0 Page 18 of 42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension 6.2 Step 2 - Develop Plant-Specific Person-Rem Dose (Population Dose)

Plant-specific release analyses were perfonned to estimate the person-rem doses to the population within a SO-mile radius from the plant. Table 5-3 provides population dose for each release category. Table 6-4 provides a correlation of CNP population dose to EPRI Accident Class. The population dose for EPRI Accident Classes 3a and 3b were calculated based on the guidance provided in EPRI Report No. 10.18243 [Reference 24] as follows:

EPRI Class 3a Population Dose= 10

• 1.01£+3 = 1.01£+4 EPRI Class 3b Population Dose= 100

• 1.01£+3 = 1.01£+5 Table 6 Population Doses Unit 2 Population Class Description Dose (person-rem)

No containment failure 1 01E+03 2 Large containment isolation failures 3.84E+06 3a Small isolation failures (liner breach) 1.01E+041 3b Large isolation failures (liner breach) 1 01E+0S2 4 Small isolation failures - failure to seal (type B) NIA 5 Small isolation failures - failure to seal (type C) NIA 6 Containment 1solabon failures (dependent failure, personnel errors) N/A 7 Severe arodent phenomena induced failure (early and late) 3.84E+06 8 Containment bypass 9.68E+06

1. 10*L.

2 100*La Revision 0 Page 19 of 42

PRA-QNT-011 Evaluation of Rlsk Significance of Short-Term ILRT Extension Table 6 Unit 2 Risk Profile for 3 in 10 Year ILRT Class Description Frequency Contribution Populatlon Population

(/yr) (%) Dose (person- Dose Rate rem) (person-rem/yr) 1 No containment fa1Iure2 2 55E-05 9349% 1 01E+03 2 58E-02 2 Large containment 1solat1on failures 1.57E-07 0.58% 3 84E+06 6.03E-01 Small isolation failures (liner 3a 2.38E-07 0 87% 1.01E+04 2.40E-03 breach)

Large 1solabon failures (liner 3b 5.92E-08 0.22% 1.01 E+05 5.98E-03 breach)

Small 1solabon failures - failure to E1 E1 E1 4 E1 seal (type B)

Small isolabon failures - failure to E1 E1 E1 5 E1 seal (type C)

Containment 1solabon failures 6 (dependent failure, personnel E1 E1 E1 E1 errors)

Severe accident phenomena 7 1.12E-06 411% 3.84E+06 4.30E+00 induced failure (early and late) 8 Containment bypass 2 00E-07 073% 9 68E+06 1.94E+00 Total 2.73E-05 6.88E+00 1 E represents a probab11isbcally 1ns1gnrticant value or a Class that Is unaffected by the Type A ILRT.

2. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total GDF.

6.3 Step 3 - Evaluate Risk Impact of Temporarily Extending Type A Test Interval from 15 to 16.5 Years The next step is to evaluate the risk impact of a one-time extension of the test interval from its current 15-year interval to a 16.5-year interval. To do this, an evaluation must first be made of the risk associated wrth the 15-year interval to allow calculation of a change in risk for the one-time extension to a 16.5-year interval.

Risk Impact Due to 15-Year Test Interval As previously stated, Type A tests impact only Class 3 sequences. For Class 3 sequences, the release magnitude is not impacted by the change in test interval (a small or large breach remains the same, even though the probability of not detecting the breach increases). Thus, only the frequency of Class 3a and Class 3b sequences is impacted. The risk contribution is changed based on the NEI guidance as described in Section 5.3 by a factor of 15/3 compared to the 3 in 1O year ILRT interval case values. The Class 3a and 3b frequencies are calculated as follows:

15 2 15 2 Frequ 2ctassJal5yr = 3

• -217

-* (CDF - (LERF - LERF1eak)) = 3

• -217

-

• 2.SBE-5 = 1.19E-6

• ~ *~

15 15 Frequ2c1ass3b15yr = 3 218

• (CDF - (LERF - LERF1eak)) = 3 218

• 2.SBE-5 = 2.96E-7 The results of the calculation for a 15-year interval are presented in Table 6-6.

Revision 0 Page 20 of 42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension Table 6 Unit 2 Risk Profile for Once in 15 Year ILRT Class Description Frequency Contribution Populatlon Population

(/yr) (%) Dose (parson- Dose Rate ram) (parson-rem/yr)

No containment fa1Iure 2 2 43E-05 89.14% 1.01E+03 2.46E-02 2 Large containment 1solat1on failures 1.57E-07 0 58% 3 84E+06 6 03E-01 Small isolation failures (!mer 3a 1.19E-06 4.36% 1.01E+04 1 20E-02 breach)

Large isolation failures (liner 3b 2 96E-07 1.08% 1.01 E+05 2 99E-02 breach)

Small isolation failures - failure to 4 £1 £1 £1 £1 seal (type B)

Small isolation failures - failure to 5 £1 £1 £1 £1 seal (type C)

Containment isolabon failures 6 (dependent failure, personnel £1 £1 £1 £1 errors)

Severe aCCJdent phenomena 7 112E-06 411% 3.84E+06 4 30E+00 induced failure (earty and late) 8 Containment bypass 2 00E-07 0.73% 9 68E+06 1.94E+00 Total 2.73E-05 6.91E+00

1. £ represents a probab1hsbcally insignificant value or a Class that IS unaffected by the Type A ILRT 2 The Class 1 frequency Is reduced by the frequency of Class 3a and Class 3b m order to preserve total GDF.

Risk Impact Due to 16.5-Year Test Interval The risk contribution for a 16.5-year interval is calculated in a manner similar to the 15-year interval. The difference is in the increase in probability of leakage in Classes 3a and 3b. For this case, the value used in the analysis is a factor of 16.5/3 compared to the 15-year interval value, as described in Section 5.3. The Class 3a and 3b frequencies are calculated as follows:

Frequ1class3a16

• syr = -16.5 3

2

• -217 * ( CDF - (LERF - LERFieak)

)

= -16.5 3

• - 2 217

• 2.SBE-5 = 1.31E-6 Frequ1class3b16 *syr = 163' 5 * ~218

• (CDF - (LERF - LERFieak)) = 163' 5 * ~

218

• 2.SBE-5 = 3.25E-7 The results of the calculation for a 16.5-year interval are presented in Table 6-7.

Revision 0 Page 21 of 42

PRA-QNT-011 Evaluation of Risk Slgntficance of Short-Tenn ILRT Extension Table 6 Unit 2 Risk Profile for Once In 16.5 Year ILRT Class Description Frequency Contribution Population Dose Population

(/yr) (%) (person-rem) Dose Rate (person-rem/yr) 1 No containment fa1Iure2 2.42E-05 88 59% 1.01 E+03 2.44E-02 Large containment lsolabon 2 1.57E-07 058% 3.84E+06 6.03E-01 failures Small 1solabon failures (liner 3a 1.31E-06 4 79% 1.01E+04 1.32E-02 breach)

Large 1solabon failures (liner 3b 3.25E-07 119% 1.01E+05 3.29E-02 breach)

Small Isolatlon failures - failure E1 E1 E1 E1 4

to seal (type 8)

Small 1solatJon failures - failure E1 E1 E1 E1 5

to ~ I (type C)

Containment !solabon failures 6 (dependent failure, personnel E1 E1 E1 E1 err~rs)

Severe acadent phenomena 7 1.12E-06 4.11% 3.84E+06 4.30E+00 induced failure (ear1y and late) 8 Containment bypass 2.00E-07 0.73% 9.68E+06 1.94E+00 Total 2 73E-05 6 92E+OO

1. E represents a probab1hstically insignificant value or a Class that IS unaffected by the Type A ILRT.

2. The Class 1 frequency Is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.

6.4 Step 4 - Determine the Change in Ri~k in Terms of LERF The risk increase associated with one-time extension of the ILRT interval involves the potential a

that a core damage event that normally would result in only small radioactive release from an intact containment could, in fact, result in a larger release due to the increase in probability of failure to detect a pre-existing leak. With strict adherence to the EPRI guidance, 100% of the Class 3b contribution would be considered LERF.

Regulatory Guide 1.174 [Reference 4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 [Reference 4] defines "very small" changes in risk as resulting in increases of CDF less than 1o-6/year and increases in LERF less than 1CJ7/year, and "small" changes in LERF as less than 10-6/year. Since containment overpressure is not required in support of ECCS performance to mitigate design basis accidents and no equipment in the shield building is credited in the CDF model at CNP, the ILRT extension does not impact CDF. Therefore, the relevant risk-impact metric is LERF.

For CNP, 100% of the frequency of Class 3b sequences can be used as a very conservative first-order estimate to approximate the potential increase in LERF from the ILRT interval extension (consistent with the EPRI guidance methodology). Based on a 15--year test interval from Table 6-6, the Class 3b frequency is 2.96E-7/year for Unit 2; based on a 16.5-year test interval from Table 6-7, the Class 3b frequency is 3.25E-7/year for Unit 2. Thus, the increase in the overall probability of LERF due to Class 3b sequences that is due to the one-time ILRT test interval extension from 15 to 16.5 years is 2.96E-8/year for Unit 2. As can be seen, even with the conservatisms included in the evaluation (per the EPRI methodology), the estimated one-Revision 0 Page 22 of 42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension time change in LERF meets the criteria for a "very small" change when comparing the 16.5-year results to the current 15-year requirement. Table 6-8 summarizes these results.

Table 6 Impact on LERF due to Extended Type A Testing Intervals ILRT Inspection Interval Unit 2: 15 Years Unit 2: 16.5 Years Class 3b (Type A LERF) 2.96E-07 3.25E-07

.6LERF (15 year baseline) 2 96E-08 NEI 94-01 [Reference 1] states that a "smalr population dose is defined as an increase of ::5 1.0 person-rem per year, or ::5 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. As shown in Table 6-9, the results of this calculation meet the dose rate criteria.

Table 6 Impact on Dose Rate due to Extended Type A Testing Intervals ILRT Inspection Interval Unit 2: 16.5 Years

.6Dose Rate (15 year baseline) 4 04E-03

%.6Dose Rate (15 year baseline) 0.0584%

1. .6Dose Rate is the difference In the total dose rate between cases. For instance, '.6Dose Rate (15 year baseline)' for the 1 in 16 5 case Is the total dose rate of the 1 In 16.5 case minus the total dose rate of the 1 m 15 year case.

2  %.6Dose Rate is the noose Rate d1V1ded by the total baseline dose rate. For instance, '%.6Dose Rate (15 year baseline)' for the 1 m 16 5 case Is the '.6D ose Rate ( 15 year baseline)' of the 1 m 16 5 year case d1V1ded by the total dose rate of the 1 m 15 year case.

6.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability Another parameter that the NRC guidance in RG 1.174 [Reference 4] states can provide input into the decision-making process is the change in the conditional containment failure probability (CCFP). The CCFP is defined as the probability of containment failure given the occurrence of an accident. This probabilrty can be expressed using the following equation:

CCFP =1_ f(ncf)

CDF where f(ncf) is the frequency of those sequences that do not result in containment failure; this frequency is detenTiined by summing the Class 1 and Class 3a results. Table 6-1 0 shows the steps and results of this calculation.

Table 6 Impact on CCFP due to Extended Type A Testing Intervals ILRT Inspection Interval Unit 2: 15 Years Unit 2: 16.5 Years f(ncf) (/yr) 2 550E-05 2.547E-05 f(ncf)/CDF 0 9350 0 9339 CCFP 0 0650 0 0661

.6CCFP (15 year baseline) j 0.108%

As stated in Section 2.0, a change in the CCFP of up to 1.5% is assumed to be "small." The one-time increase in the CCFP from the 1 in 15 year interval to 1 in 16.5 year interval is 0.108%

for Unit 2. Therefore, this increase is judged to be "small."

Revision 0 Page 23 of 42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension 6.6 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage An estimate of the likelihood and risk implications of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is evaluated using a methodology similar to the Calvert Cliffs liner corrosion analysis [Reference 5]. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete basemat, each with a steel liner.

The following approach is used to determine the change in likelihood, due to extending the ILRT, of detecting corrosion of the containment steel liner. This likelihood is then used to determine the resulting change in risk. Consistent with the Calvert Cliffs analysis, the following issues are addressed:

Differences between the containment basemat and the containment cylinder and dome The historical steel liner flaw likelihood due to concealed corrosion The impact of aging The corrosion leakage dependency on containment pressure The likelihood that visual inspections will be effective at detecting a flaw Assumptions Consistent with the Calvert Cliffs analysis, a half failure is assumed for basemat concealed liner corrosion due to the lack of identified failures (See Table 6-11, Step 1).

In the 5.5 years following September 1996 when 10 CFR 50.55a started requiring visual inspection, there were three events where a through wall hole in the containment liner was identified. These are Brunswick 2 on 4/27/99, North Anna 2 on 9/23/99, and D. C.

Cook 2 in November 1999. The corrosion associated with the Brunswick event is believed to have started from the coated side of the containment liner. Although CNP has a different containment type than Brunswick 2 and North Anna 2, these events could also potentially occur at CNP (i.e., corrosion starting on the coated side of containment).

Construction material embedded in the concrete may have contributed to the corrosion.

The corrosion at North Anna is believed to have started on the uninspectable side of containment due to wood embedded in the concrete during construction. The D.C. Cook event is associated with an inadequate repair of a hole drilled through the liner during construction. Since the hole was created during construction and not caused by corrosion, this event does not apply to this analysis. Based on the above data, there are two corrosion events from the 5.5 years that apply to CNP.

Consistent with the Calvert Cliffs analysis, the estimated historical flaw probability is also limited to 5.5 years to reflect the years since September 1996 when 10 CFR 50.55a started requiring visual inspection. Additional success data was not used to limit the aging impact of this corrosion issue, even though inspections were being performed prior to this date (and have been performed since the time frame of the Calvert Cliffs analysis)

(See Table 6-11, Step 1).

Consistent with the Calvert Cliffs analysis, the steel liner flaw likelihood is assumed to double every five years. This is based solely on judgment and is included in this analysis to address the increased likelihood of corrosion as the steel liner ages (See Table 6-11, Steps 2 and 3). Sensitivity studies are included that address doubling this rate every ten years and every two years.

In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reaching the outside atmosphere, given that a liner flaw exists, was estimated as 1.1 % for the cylinder and dome, and 0.11 % (10% of the cylinder failure probability) for the basemat. These Revision 0 Page 24 of 42

PRA-QNT-011 Evaluatlon of Risk Significance of Short-Tenn ILRT Extension values were determined from an assessment of the probability versus containment pressure. For CNP, the ILRT maximum pressure Is 12.0 psig [Reference 30].

Probabilities of 1% for the cylinder and dome, and 0.1 % for the basemat are used in this analysis.

Consistent with the Calvert Cliffs analysis, the likelihood of leakage escape (due to crack formation) in the basemat region is considered to be less likely than the containment cylinder and dome region (See Table 6-11, Step 4).

In the Calvert Cliffs analysis, It is noted that approximately 85% of the interior wall surface is accessible for visual inspections. Consistent with the Calvert Cliffs analysis, a 5% visual inspection detection failure likelihood given the flaw is visible and a total detection failure likelihood of 10% is used. To date, all liner corrosion events have been detected through visual inspection (See Table 6-11, Step 5).

Consistent with the Calvert Cliffs analysis, all non-detectable containment failures are assumed to result in early releases. This approach avoids a detailed analysis of containment failure timing and operator recovery actions.

Revision 0 Page 25 of 42

PRA-QNT --011 Evaluation of Risk Significance of Short-Tenn ILRT Extension Table 6 Steel Liner Corrosion Base Case Step Description Containment Cylinder and Containment Basemat (15%)

Dome (85%)

H1stoncal liner flaw likelihood Events* 2 Events 0 Failure data containment locabon (Brunswick 2 and North Anna 2) Assume a half failure speCJfic 2 / (70 X 5.5) = 5.19E-03 0.5 / (70 X 5.5) = 1.30E-03 Success data* based on 70 steel-lined containments and 5.5 years since the 10 CFR 50.55a requirements of periodic VJsual inspections of containment surfaces Aged adjusted liner flaw likelihood Year Failure rate Year Failure rate Dunng the 16.5-year interval, 1 2.05E-03 1 5.13E-04 assume failure rate doubles every 16 5 1.76E-02 16.5 4 41 E-03 2 five years (14.9% increase per year). The average for the 5th to 10th year set to the histoncal failure 16.5 year average= 8 34E-03 16.5 year average = 2.09E-03 rate.

Increase In flaw likelihood between 3 and 15 years Uses aged adjusted 0 71% (1 to 3 years) 0.18% (1 to 3 years) 3 liner flaw likelihood (Step 2), 9 66% (1 to 15 years) 2 42% (1 to 15 years) assuming failure rate doubles every 12 52% (1 to 16.5 years) 3.13% (1 to 16.5 years) five years.

Likelihood of breach In containment 4 1% 0.1%

given liner flaw 10%

5% failure to 1dentrt'y VJsual flaws plus 5% likelihood that the flaw is not visible (not through-cylinder Visual inspection detection failure 100%

5 but could be detected by ILRT).

likelihood Cannot be visually inspected All events have been detected through visual inspection. 5%

v1sIble failure detecbon Is a conservative assumption.

0.00071% (3 years) 0 00018% (3 years) 0.71%x1%x10% 0 18% X 0.1 % X 100%

Ukel1hood of non-detected 0.00966% (15 years) 0.00242% (15 years) 6 containment leakage (Steps 3 x 4 x

5) 9.66% X 1% X 10% 2.42% XO 1% X 100%

0.01252% (16.5 years) 0.00313% (16.5 years) 12.52% X 1% X 10% 3.13% XO 1% X 100%

The total likelihood of the corrosion-induced, non-detected containment leakage is the sum of Step 6 for the containment cylinder and dome, and the containment basemat, as summarized below for CNP.

Table 6-12 -Total Likelihood on Non-Detected Containment Leakage Due to Corrosion for CNP Description At 3 years 0 00071 % + o 00018% = 0.00089%

At 15 years 0.00966% + 0.00242% = 0.01207%

At 16.5 years 0.01252% + 0 00313% = 0.01565%

Revision 0 Page 26 of 42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension The above factors are applied to those core damage accidents that are not already independently LERF or that could never result in LERF.

The two corrosion events that were initiated from the non-visible (backside) portion of the containment liner used to estimate the liner flaw probability in the Calvert Cliffs analysis are assumed to be applicable to this containment analysis. These events, one at North Anna Unit 2 (September 1999) caused by timber embedded in the concrete immediately behind the containment liner, and one at Brunswick Unit 2 (April 1999) caused by a cloth work glove embedded in the concrete next to the liner, were initiated from the nonvisible (backside) portion of the containment liner. A search of the NRC website LER database identified two additional events have occurred since the Calvert Cliffs analysis was performed. In January 2000, a 3/16-inch cir~ular through-liner hole was found at Cook Nuclear Plant Unit 2 caused by a wooden brush handle embedded immediately behind the containment liner. The other event occurred in April 2009, where a through-liner hole approximately 3/8-inch by 1-inch in size was identified in the Beaver Valley Power Station Unit 2 (BVPS-1) containment liner caused by pitting originating from the concrete side due to a piece of wood that was left behind during the original construction tbat came in contact with the steel liner [Reference 34]. Two other containment liner through-wall hole events occurred at Turkey Point Units 3 and 4 in October 201 O and November 2006, respectively. However, these events originated from the visible side caused by the failure of the coating system, which was not designed for periodic immersion service, and are not considered to be applicable to this analysis. More recently, in October 2013, some through-wall containment liner holes were identified at BVPS-1, with a combined total area of approximately 0.395 square inches. The cause of these through-wall liner holes was attributed to corrosion originating from the outside concrete surface due to the presence of rayon fiber foreign material that was left behind during the original construction and was contacting the steel liner. For risk evaluation purposes, these five total corrosion events occurring in 66 operating plants with steel containment liners over a 17 .1 year period from September 1996 to October 4, 2013 (i.e., 5/(66*17.1) = 4.43E-03) are bounded by the estimated historical flaw probability based on the two events in the 5.5 year period of the Calvert Cliffs analysis (i.e.,

2/(70*5.5) = 5.19E-03) incorporated in the EPRI guidance [Reference 34].

6.7 Impact from External Events Contribution An assessment of the impact of external events is performed, The purpose for this investigation is to determine the change in LERF associated with a one-time 18-month extension in ILRT testing.

The CNP Fire PRA model was created to satisfy the ASME/ANS PRA standard [Reference 37]

and support risk-informed NFPA 805 applications [Reference 31]. The Fire PRA model was used to obtain the fire CDF and LERF values [Reference 28]. To reduce conservatism in the ILRT extension analysis, the methodology of subtracting existing LERF from CDF (except the risk associated with a pre-existing leak) is also applied to the Fire PRA model. The following shows the calculation for Class 3b:

15 Frequ2c1ass3b15yr =3

• Pclass3b * ( CDF - (LERF - LERF1ea1c))

= 315

• 218 0.5

• (6.81E-S - (3.70E 3.0BE-7)) = 7.42E-7 Frequ2class3b16.5yr = -16.S 3

(

-

• Pc1ass3b

• CDF - (LERF - LERFieak)

)

= -16.5 3

0.5 218

• (6.81E (3.70E 3.0BE-7)) = 8.16E-7 Revision 0 Page 27 of 42

PRA-QNT-011 Evaluation of Risk Significance of Short-Term ILRT Extension Reference 27 was developed for risk-informed applications and provides an assessment of the seismic hazard. The Unit 2 Seismic GDF and LERF are estimated to be 2.38E-5/yr and 5.36E-6/yr, respectively.

Subtracting seismic LERF from GDF (except the risk associated with a pre-existing leak), the Class 3b frequency can be calculated by the following formulas:

15 Freqclass3b15yr = 3

• Pc1ass3b * (CDF - (LERF - LERF1ea1J)

= 315

• 218 0.5

• (2.38£ (5.36£ 3.84£-8)) = 2.12E-7 16.5 Freqclass3b16 5yr =

• Pc1ass3b * ( CDF - (LERF - LERFieak)

)

= 163*5 * ~

218

• (2.38£ (5.36£ 3.84£-8)) = 2.33E-7 Reference 32 provides a high winds and tornado assessment, along with other external hazards. Due to the low frequency of strong wind, tornado, and tornado-induced missile events and the high level of protection afforded the CNP to these events, it is concluded that the contribution to plant risk from severe wind events is insignificant [Reference 32]. Similarly, it is concluded that the contribution to plant risk from external flooding, transportation accidents, other external events is insignificant [Reference 32].

The external event contributions to Class 3b frequencies are then combined to obtain the total external event contribution to Class 3b frequencies. The LERF increase is conservatively assumed to be the change in Class 3b frequency.

Table 6 Unit 2 CNP External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 1 per 15 years to 1 1 per 15 year 1 per 16.5 years per 16.5 years)

External Events 9.53E-07 1.0SE-06 9.53E-08 Internal Events 2 96E-07 3.25E-07 2 96E-08 Combined 1 25E-06 1.37E-06 1.25E-07 The internal event results are also provided to allow a composite value to be defined. When both the internal and external event contributions are combined, the increase due to temporarily increasing the interval from 15 to 16.5 years is 1.25E-7 for Unit 2, which meets the guidance for "small" change in risk, as it exceeds 1.0E-7/yr and remains less than a 1.0E-6 change in LERF

[Reference 4]. For this change in LERF to be acceptable, total LERF must be less than 1.0E-5.

The total LERF is calculated below:

LERF = LERFmternal + LERFflre + LERF.eisnuc + LERFc1ass3Bmcrea.se LERF16.syr = l.78E-6/yr + 3.70E-6/yr + 5.36E-6/yr + l.25E-7 /yr= 1.lOE-5/yr Appendix B demonstrates that sufficient conservatisrhs exist in the CNP PRA models to conclude a realistic estimate of total site LERF would remain less than 1E-5 /yr. As specified in Regulatory Guide 1.174 [Reference 4], since the total LERF is less than 1.0E-05, it is acceptable for the ilLERF to be between 1.0E-07 and 1.0E-06.

Revision 0 Page 28 of 42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension 7.0 RESULTS The results from this ILRT extension risk assessment for CNP are summarized in Table 7-1 for Unit 2.

Table 7 Unit 2 ILRT Extension Summary Class Dose 1 In 16 Years Extend to (person- 1 in 16.5 Years rem)

CDFNear Person- CDFNear Person-RemNear RemNear 1.01E+03 2.43E-05 2.46E-02 2.42E-05 2.44E-02 2 3.84E+06 1.57E-07 6.03E-01 1.57E-07 6.03E-01 3a 1.01E+04 1.19E-06 1 20E-02 1.31 E-06 1.32E-02 3b 1.01E+05 2 96E-07 2 99E-02 3.25E-07 3.29E-02 7 3.84E+06 1.12E-06 4.30E+OO 1 12E-06 4.30E+O0 8 9.68E+06 2.00E-07 1.94E+00 2.00E-07 1.94E+00 Total 2.73E-05 6.91E+00 2.73E-05 6.92E+00 ILRT Dose Rate from 3a and 3b LiTotal From 15 Dose N/A 4.04E-03 Years Rate

%LiDose From 15 NIA 0.0584%

Rate Years I

From 15 LiLERF NIA 2.96E-08 Years From 15 LiCCFP% N/A 0108%

Years Revision 0 Page 29 of 42

PRA-QNT-011 Evaluation of Risk Significance of Short-Term ILRT Extension

8.0 CONCLUSION

S AND RECOMMENDATIONS Based on the results from Section 6.0, the following conclusions regarding the assessment of the plant risk are associated with a one-time extension of the Type A ILRT test frequency to 16.5 years:

Regulatory Guide 1.174 [Reference 4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines "very small" changes in risk as resulting in increases of CDF less than 1.0E-06/year and increases in LERF less than 1.0E-07/year. Since the ILRT does not impact CDF, the relevant criterion is LERF. The one-time increase in LERF resulting from a change in the Type A ILRT test interval from 1 in 15 years to 1 in 16.5 years is estimated as 2.96E-8/year for Unit 2 using the EPRI guidance; this value increases negligibly if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included. Therefore, the estimated change in LERF is determined to be "very small" using the acceptance guidelines of Regulatory Guide 1.174 [Reference 4].

When external event risk is included, the one-time increase in LERF resulting from a change in the Type A ILRT test interval from 1 in 15 years to 1 in 16.5 years is estimated as 1.25E-7/year for Unit 2 using the EPRI guidance. The total LERF is estimated to be 1.1 0E-5/yr. Appendix B demonstrates that sufficient conservatisms exist in the CNP PRA models to conclude a realistic estimate of total site LERF would remain less than 1E-5

/yr. As such, the estimated change in LERF is determined to be "small" using the acceptance guidelines of Regulatory Guide 1.174 [Reference 4].

The effect resulting from temporarily changing the Type A test frequency to 1-per-16.5 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing is 0.0040 person-rem/year for Unit 2. NEI 94-01

[Reference 1] states that a "small" population dose is defined as an increase of :5 1.0 person-rem per year, ors 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. The results of this calculation meet these criteria. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible.

The one-time increase in the conditional containment failure probability from the 1 in 15 year interval to 1 in 16.5 year interval is 0.11 % for Unit 2. NEI 94-01 [Reference 1] states that increases in CCFP of s 1.5% is "small." Therefore, this increase is judged to be "small."

Therefore, the one-time ILRT interval extension to 16.5 years is considered to be "small" since it represents a small change to the CNP risk profile.

Previous Assessments The NRC in NUREG-1493 [Reference 6] has previously concluded that:

Reducing the frequency of Type A tests (ILRTs) from 3 per 10 years to 1 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.

Revision 0 Page 30 of42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension 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 1 in 20 years has not been evaluated. Beyond testing the performance of containment penetrations, ILRTs also test integrity of the containment structure.

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

Revision 0 Page 31 of 42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension A. PRA TECHNICAL ADEQUACY FOR ILRT The CNP Internal Events PRA (including internal flooding) received a full scope peer review in 2015, followed by several focused-scope reviews on various portions of the model, such as pre-initiator HRA and containment hydrogen analysis [Reference 17]. For the purposes of this ILRT extension evaluation, which only requires an assessment of CC-I, only those SRs that are currently "Not Met" are evaluated. Of the remaining Internal Events SRs, no impacts were identified on the ILRT extension evaluation.

The CNP Fire PRA [Reference 28] was subject to a full scope peer review during initial model development in 2010 [Reference 43], with follow-on focused-scope reviews occurring in 2015

[Reference 44] and 2017 [Reference 45]. For the purposes of ILRT extension evaluation, which only requires an assessment of CC-I, only those SRs that are currently "Not Met" are evaluated.

Of the remaining Fire PRA SRs, open items related to IGN-A 1 are identified as a potential impact on the ILRT extension evaluation. This SR is evaluated as Not Met due to the use of fire ignition frequencies based on NUREG/CR-6850 Supplement 1 [Reference 33] instead of NUREG-2169 [Reference 42]. For the purposes of the ILRT extension evaluation, the Fire PRA model is quantified with each set of fire ignition frequencies, and the more limiting results are used.

The CNP Seismic PRA peer review was conducted in 2018 [Reference 46], with a formal F&O closure review in 2019 [Reference 47]. For the purposes of ILRT, which only requires an assessment of CC-I, only those SRs that are "Not Met' are listed and evaluated. However, no SPRA-related SRs are currently "Not Met" [Reference 48], so no evaluation is provided for the purposes of the ILRT extension evaluation.

Revision 0 Page 32 of42

PRA-QNT -011 Evaluatlon of Risk Significance of Short-Tenn ILRT Extension B. TOTAL SITE RISK DISCUSSION 8.1. Purpose The total LERF is estimated in Section 6. 7 to be 1.1 0E-5/yr. The purpose of this attachment is to demonstrate that sufficient conservatisms exist in the DC Cook PRA models to conclude that a realistic estimate of total site LERF would remain below 1E-5 /yr. These conservatisms are discussed both qualitatively and quantitatively in this appendix.

8.2. Identification of Model Conservatisms The DC Cook LERF model contains several quantitative and qualitative conservatisms that are discussed in this section. The Seismic PRA (SPRA) contributes nearly half the total LERF, primarily due to the risk importance of the Distributed Ignition System (DIS). The table below discusses some of the identified conservatisms, along with a qualitative discussion on the expected impact to the LERF result.

Of each of the modeled hazards, the SPRA contributes the most to the total, at nearly 50%. The Fire PRA contributes the second highest at 34%, with the FPIE model contributing the remainder. The SPRA is a newer model, with significant conservatisms in the fragility modeling and is considered to be much less realistic than either the FPIE or Fire PRA models.

Revision 0 Page 33 of 42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension Table B Qualitative LERF Model Conservatisms Model(s) Expected Conservatism Discussion Impacted Benefit to LERF The SPRA LERF estimate IS the largest contnbutor to the total LERF, contnbuting nearly half of the total. This IS in part due to a lack of independent power to DIS and also due to the conserva!Jve nature of eslJmalJng seismic frag1lrt1es of components. Like components are assumed to be fully correlated (i e fail simultaneously m all seismic events where their corresponding fragilrty Seismic events probab1hs1Jcally fa1Q The DC Cook Se1sm1c PRA also s1mphf1es the SPRA Moderate to High Frag1lltles assessment of when the seismic frag1lrty failure mode actually fails the component The fragility evaluation assumes that the lowest capacrty failure mode, typically an anchorage failure, fails the component While this may be the case for complex active components like EDGs, for more passive comoonents such as heat exchanaers this is verv conservalive.

DC Cook 1s an ice condenser containment design, which 1s smaller than other PWR containment designs and has DIS 1gnlters instaDed to control containment hydrogen post-acadent S1gmflcant LERF benefit IS obtained m the Se1sm1c PRA If DIS is recovered pnor to significant hydrogen buildup m containment DC Cook 1s currently pursuing a permanent modrflcalJon to provide backup power to DIS If onsrte emergency power IS lost However, the eXJstmg FLEX plan for DC Cook provides for re-energ1zmg the lgnrters at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> post event.

While this Is too late to provide benefrt for core damage sequences that occur Distnbuted early In the event, longer developing core damage sequences would benefit lgnrtJon from credrtmg this FLEX strategy.

SPRA Moderate System (DIS)

Recovery The SPRA MOR shows the Fussell-Vesely Importance of the Loss of Offslte Power sequences make up approximately 9% of the total LERF, much of which can be mibgated by FLEX If DIS recovery is also credited. The existing documented FLEX sens1t1vity did not show significant reduction in LERF in part because DIS recovery was not credrted for the SPRA sensrtMly.

Loss of offs1te power events at DC Cook result In core damage pnmanly due to RCP seal LOCAs, Se1sm1cally-lnduced Very Small LOCAs, or loss of feedwater The LOCA core damage sequences tend to be long developing loss of inventory sequences wrth suffiCJent time available to energize DIS from Revision o Page 34 of42

PRA-QNT-011 Evaluation of Risk Significance of Short-Term ILRT Extension Conseivatlsm Model(s) Expected Discussion lmoacted Benefit to LERF FLEX. Loss of feedwater sequences tend to result in core damage rapidly since the TDAFP battenes only last 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> However, more detailed a-edit for deep load shed and/or manual TDAFP operallon as planned for FLEX would reduce the LERF lmoact from these seauences FLEX is modeled in the updated 2020 DC Cook FPIE Model but IS set to failed for this nsk assessment since rt is considered a PRA upgrade that has not yet been peer reviewed DC Cook's FLEX strategy provides feedwater and small RCS Inventory makeup, along with re-energ1zmg key instruments and DIS (see DIS Recovery for a d1Scuss1on of this strategy) If modeled, FLEX reduces the FPIE Unrt 2 LERF from 1.78E-6 to 1.64E-6.

Smee the FLEX model already exists m the FPIE model, rt Is adapted for the Fire PRA model of record for the purposes of sensItIvrty cases to demonstrate FPIE, Fire FLEX Moderate the potential nsk reduction m the Fire PRA This numenc sensrtivlty Is PRA, SPRA discussed later In thlS section The SPRA contairis a sunpllfied version of the updated FLEX modeling that Is currently modeled (but inactive for this nsk assessment) m the FPIE model.

The updated modeling expands FLEX credit to smaller RCP seal LOCAs and small LOCAs not credited m the SPRA MOR, since thermal-hydraulic analyses demonstrating successful preven!Jon of core damage were not available when the SPRA was comoleted EPRI 1018243, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, counts EPRI Class 3b (Large pre-existing leak in containment) frequency as part of the delta LERF es!Jmates for the assessment This frequency is also included m the base PRA model as a pre-EPRI Accident FPIE, Fire exIs!Jng containment bypass. EPRI notes In the report that no pre-existmg Low to Moderate Class 3b PRA, SPRA leaks have been detected m the industry that would consbtute a large release The frequency Is calculated using a Jefferey's non-mforrna!Jve pnor method, which Is conservabve A more reailsllc assessment of this frequency would v10ld better LERF es!Jmates.

Medium LOCA Containment Bypass sequences (those sequences m which the core damage Containment sequence also bypasses containment) are the most nsk-significant LERF Moderate (FPIE),

FPIE, SPRA Bypass contributor to the FPIE model, contnbullng approXJmately 40% of total LERF. Low (SPRA)

Revision 0 Page 35 of 42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension Model(s) Expected Conservatism Discussion Impacted Benefit to LERF Of these, Medium LOCA contnbutes nearly 30% of LERF sequences that result from failure of ECCS. These sequences are assumed to result in LERF due to potential containment overpressure from failure of core cooling. However, containment spray JS not credited since 1! would not prevent core damage.

Containment spray has separate heat exchangers and pumps that can take suction directly from the containment sump and operates mostly independently of RHR (although there are some shared isolation valves). If containment spray was credited, containment overpressure would not necessanly be assured in these sequences. Other subsequent failures, such as failure of DIS or induced tube rupture, would still potentially result in LERF for these sequences but would not be guaranteed Some moderate reduction in the LERF estimate would be exoected rf more realistic modeling were adopted A study of DC Cook's containment overpressure indicates the likely failure location Is at the basemat, due to diagonal tension failure where the containment cylinder connects to the basemat (PRA-L2-Model, Revision 2, Attachment 1, Page 132). The HCLPF of this failure mechanism Is assumed to be the representative HCLPF for containment failure in the updated DC Cook Containment Hydrogen Overpressure analysis. Depending on the accident Containment sequence, this location may be partially or completely underwater, which may FPIE, Fire Failure result In some rad1onuchde scrubbing. For simphcrty, overpressure failure of Low to Moderate PRA, SPRA Location containment Is assumed to always result in a Large Early Release to account for the uncertainty with respect to the exact failure location Due to this uncertainty, no quantitative prediction can be made as to what impact this would have on LERF. However, 111s reasonable to conclude that rf a more realistic treatment were applied, some reduction In the LERF estimate would occur.

NUREG-2178 Volume 2 contains updated heat release rates for electric motors NUREG-2178 and indoor dry transformers, which have not been incorporated into the Fire Volume 2- PRA While these are not the most risk s1gnrflcant fires at DC Cook, a low to Fire PRA Low to Moderate Updated Heat moderate reduction in Fire RJSk estimates would likely be seen If some target Release Rates damage Is removed, particularly for transformers In the electncal switchgear areas.

Revl51on O Page 38 of42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension Conservatism Discussion Model(s) Expected Impacted Benefit to LERF NUREG-2178 Volume 2 provides an altemabve approach to modeling fires from the mam control boards to the AppendIX L method m NUREG/CR-6850.

NUREG-2178 The AppendIX L method ts still In use at DC Cook. The NUREG-2178 method Is Volume 2 -

a more realistic treabnent of the mam control board fires and Is expected to Mam Control reduce the risk from these fires. Fire scenanos m the Mam Control Room Fire PRA Low Board (MCB)

(MCR) make up approximately 4% of LERF m the current Fire PRA Model of Fire Scenano Record (MOR), and MCB fires are generally more important to NFPA 805 delta Event Tree nsk than they are to total nsk. For this reason, thJS method is likely to only provide a low benefit to thJS application NUREG-2230 provides updated detection and suppression event trees for electJncal cabinet fires, along with new estimates for mean suppression rates.

Electncal cabinet fires are very significant nsk contributors at DC Cook, making NUREG-2230 up several of the most risk s1gnlf1cant fire scenarios for LERF. The risk Suppression s19niflcant panels tend to be switchgear panels m the main electrical switchgear and Detection Fire PRA Moderate rooms, which are adjacent to opposite tram panels with numerous cables for Electncal running In overhead raceways. This fire area Is protected by an automatic C.2 Cabinet Fires suppression system and would likely see moderate benefit to the nsk estimate by elImInating some cable tray damage from electrical switchgear fires in this area NUREG-2233 DRAFT NUREG-2233 provides updated fire modeling inputs (heat release Likely Moderate, (DRAFT) rates, zones of influence) for transient fires While not yet approved, many nsk Uncertain since Updated Fire PRA s1gnlf1cant fires m the DC Cook Fire PRA are transient fires and as such, some method Is not Transient Fire potential reduction of nsk estimates Is expected when the method Is finalized final Methodoloqy Earty The 2020 FPIE MOR contains updated Early Containment Failure Probablllties Containment (CFEs) compared to the 2017 Fire PRA MOR. These values were updated sl19htly post peer review to more realistic values and are included as part of the Fire PRA Low Failure Probab1lrtJes FLEX sensrtivrty.

Revision O Page 37 of 42

PRA-QNT-011 Evaluation of Risk Significance of Short-Term ILRT Extension Model(s) Expected Conservatism Discussion Impacted Benefit to LERF Radionuclide scrubbing may be possible If containment spray operates, rt may also be possible in certain SGTR and ISLOCA sequences that bypass containment if water IS in the release pathway Scrubbing Is not credited due to Credit for the drfflculty in identifying the specific sequences in which rt wouk:! be Rad1onud1de applicable However, some benefit to the total LERF estimate would be FPIE Low Scrubbing expected If a more realistic treatment were adopted, particularly for SGTR sequences where feedwater JS available to the ruptured SG or ISLOCA events WCAP-16341-P discusses the JSSue of contaInment spray scrubbing in Secl!on 3 4 and the complexrt:Jes surroundinq credrtinA 11 for lhJS purpose.

Revision O Page 38 of 42

PRA-QNT-011 Evaluatlon of Risk Significance of Short-Tenn ILRT Extension 8.2.1 FLEX Quantitative Sensitivity Development This year, the DC Cook FPIE Model update included a revised, detailed FLEX model. This FLEX model is based on.

The HRA guidance in EPRI 3002013018, "Human Reliability Analysis (HRA) for Diverse and Flexible Mitigation Strategies (FLEX) and Use of Portable Equipment' The implementation guidance in PWROG-14003-NP, Implementation of FLEX Equipment in Plant-Specific PRA Models, Revision 2, August 2016 The portable equipment failure rates in PWROG-18043-P, FLEX Equipment Data Collection and Analysis, Revision 0, February 2020 This sensitivity was performed by modifying the 2017 Fire PRA MOR using the fault tree logic from the 2020 FPIE MOR. The updated FLEX model is considered a PRA upgrade, and is not credited for the base case results in this analysis. The FPIE FLEX model is adapted for a sensitivity in the Fire PRA MOR to estimate the impact of FLEX in the Fire PRA MOR LERF value.

The following gates were copied over from the 2020 FPIE MOR, with destination fault trees listed in parentheses. Existing logic in the destination fault trees is modified as necessary to accommodate this logic (such as connecting to the same top gate dependencies developed for the FPIE mode). These changes primarily include retaining credit for the Fire PRA manual TDAFP operation and ESW supply credit to AFW.

G2-FLEX-FW-P1 (2af cal)

G2-FLEX-FW-P2 (2af caf)

• G2-FLX-EZC-B-P2 (2mcc caf)

G-2AABS-21DFALBE, G-2ABBS-21BFALBE, G-2ABMC-EZCBFAL-BE, G-1ABBS-11 BFALBE, G-1AABS-11 DFALBE (3FIRE-EXCLUS/ONS ca()

• G2-FLX-DCB-P1 (2dcb caf)

• G2-FLX-DCB-P2 (2dcb caf)

G2-FLX-CRID4-P1 (2p4.caf')

G2-FLX-CRID4-P2 (2p4 caf)

G2-FLEX-CDWN-3HR-P1 (2oa6 ca()

• G2-FLEX-SGPORV-P1 (2ms ca()

• G2-FLX-RCSINV-P2 (2hpl cafJ G2DN-FLEX-P1 (2dcn ca()

• G2DN-FLEX-P2 (2dcn ca()

G1-FLX-11 D-P2 (11D caf)

• G1-FLX-CRID4-P1 (1p4.caf')

• G1-FLX-CRID4-P2 (1p4.caf)

• G2-FLX-21D-P2 (21D.caf)

G2-FLX-21 B-P2 (21 B caf)

Revision 0 Page 39 of 42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension The updated Fire-Induced station Blackout (FSBO) event tree is also imported from the FPIE model. Updates to the top gate connection logic (2in-fire.caf) and the template file for two top creation (DCC-Template-Fire.caf) are completed as necessary to accommodate the new sequence logic.

Wrth these changes completed, the model is rebuilt and quantified using the same process used to build the Fire PRA MOR.

In addition to the FLEX sensitivity, the updated CFE values are included as a separate sensitivity on top of the FLEX sensitivity to determine its impact on the total LERF estimates.

This sensitivity is quantified in the same manner as the Fire PRA MOR.

The purpose of these sensitivities is to understand the impact of modeling updates and conservatisms on the total LERF risk estimates. Due to the way these sensitivities are performed, several issues arise, such as mixing of older and newer reliability data, and the lack of updated HRA dependency between existing Fire HFEs and the FLEX HFEs. Given that the FLEX failures typically go directly to core damage due to lack of feedwater or inventory control, it is not expected this will significantly impact the result and is judged to be adequate for this sensitivity.

The results of these sensitivities are shown in the tables below.

Revision 0 Page 40 of42

PRA-QNT-011 Evaluation of Risk Significance of Short-Tenn ILRT Extension 8.2.2 FLEX Quantitative Sensitivity Results Table B FLEX & CFE Senslbvlty Results Case Cll98 D99crlpbon FLEX CFE Update Unlt2 LERF Credit (FPRA)

Base + Old CFE No No 3.70E-06 2 Base + New CFE No Yes 3.64E-06 3 FLEX + Old CFE Yes No 3.03E-06 4 FLEX + New CFE Yes Yes 2 98E-06 Table 8 Total LERF Estimates for Sensitivity Caae11 Case Unlt2 LERF Unlt2 LERF Un,t2 LERF Total LERF Total LERF (FPIE

• No FLEX) (FPIE

• FLEX) (SPRA) No FPIE FLEX Crecllt FPIE FLEX Credit 1.78E-06 1 64E-06 5.36E-06 1 0BE-05 1.07E-05 2 1.78E-06 1 64E-06 5.36E-06 1 0BE-05 1.06E-05 3 1.78E-OB 1 64E-06 5.36E-06 1.02E-05 1.00E-05 4 1.78E-06 1 64E-06 5.36E-06 1.01 E-05 9.99E-06 Revision 0 Page 41 of 42

PRA-QNT--011 Evaluation of Risk Significance of Short-Term ILRT Extension B.2.3 Concluslons The sensitivity cases, combined with the qualitative discussion of modeling conservatisms, demonstrate that there is reasonable assurance that the total site LERF estimate is less than 1E-5 /yr. Case 4 shows that the overall decrease from crediting FLEX and the updated CFE values results in an estimate just less than 1E-5 /yr total LERF. While this estimate is rough, if additional qualitative conservatisms were removed, the LERF estimate would be significantly reduced.

Another issue arises when all the DC Cook Hazard PRA model risk estimates are summed to detem,ine the total site LERF estimate. PWROG-16034-NP, Implementation Guidance for Risk Aggregation Pilot, discusses this issue in detail. The primary concern with summing risk estimates in this way is that the PRA model for that hazard may be significantly more conservative than more refined models, such as the FPIE model. This can have the effect of over-inflating the risk importance of the conservative hazard with respect to the others. In the case of the DC Cook total risk estimate, it is clear the SPRA LERF estimate is conservative due the inherent conservatism in fragility evaluations and lack of refined FLEX credit. Conservatism exists in the Fire PRA as well but is less pronounced due to the additional time Fire PRA models have been in general use.

For this application, the calculational methodology to detem,ine the delta LERF estimate is straightforward, with one failure mode of concern: the potential for a large pre-existing leak to go undetected. Since the estimate is focused on the single contributor, which is modeled already in the base model, the risk importance of other equipment remains the same as it does in the base model. For this reason, this discussion focused mainly on model conservatisms inherent to the base PRA hazard models and not on any new contributors as would arise in other applications.

Overall, sufficient conservatisms exist to conclude that the total LERF estimate would remain less than 1E-5 /yr.

Revision 0 Page 42 of 42

Enclosure 4 to AEP-NRC-2020-72 Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages Marked to Show Proposed Changes

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Containment Leakage Rate Testing Program

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Perfonnance-Based Option of 10 CFR 50, Appendix J," dated July 2012, and Section 4.1, "Limitations and Conditions for NEI TR 94-01, Revision 2," of the NRG Safe Evaluation Re rt in NEI 94-01, Revision 2-A, dated October 200 exce t that the next T e est erformed after the A ril 22 2006 T e A test shall be

b. The containment design pressure is 12 psig. For the Containment Leakage Rate Testing Program, Pa is 12.0 psig.

c. The maximum allowable containment leakage rate, La, at Pa, shall be 0.18%

of containment air weight per day.

d. Leakage rate acceptance criteria are:

1. 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 s 0.60 La for the Type B and C tests ands 0.75 La for Type A tests.

2. Air lock testing acceptance criterion is overall air lock leakage rate is s 0.05 La when tested at~ Pa.

e. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

5.5.15 Battery Monitoring and Maintenance Program This program provides for battery restoration and maintenance, based on the recommendations of IEEE Standard 450-2010, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," as endorsed, with certain regulatory positions, in Regulatory Guide 1.129, Revision 3, or of the battery manufacturer including the following:

a. Actions to restore battery cells with float voltage < 2.13 V; and

b. Actions to equalize and test battery cells that had been discovered with electrolyte level below the minimum established design limit.

Cook Nuclear Plant Unit 2 5.5-14 Amendment No. 2-78, 600, a.+4-, ~ , 325