ML15072A264

Issuance of Amendments Containment Leakage Rate Testing Program (TACs MF3568 and MF3569)

ML15072A264
Person / Time
Site:	Cook
Issue date:	03/30/2015
From:	Mahesh Chawla, Dietrich A Plant Licensing Branch III
To:	Weber L Indiana Michigan Power Co
Chawla M
References
TAC MF3568, TAC MF3569
Download: ML15072A264 (35)

Text

Mr. Lawrence J. Weber Senior Vice President and Chief Nuclear Officer UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 30, 2015 Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, Ml 49106

SUBJECT:

DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: CONTAINMENT LEAKAGE RATE TESTING PROGRAM (TAC NOS. MF3568 AND MF3569)

Dear Mr. Weber:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 326 to Renewed Facility Operating License No. DPR-58 and Amendment No. 309 to Renewed Facility Operating License No. DPR-74 for the Donald C. Cook Nuclear Plant (CNP), Units 1 and 2 respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated March 7, 2014, as supplemented by letters dated September 30, 2014, December 16, 2014, January 15, 2015, and February 20, 2015.

The amendments revise the CNP Units 1 and 2, TS 5.5.14, "Containment Leakage Rate Testing Program," by adopting Nuclear Energy Institute (NEI) 94-01 Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J" (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12221A202), and Section 4.1, "Limitations and Conditions for NEI [Topical Report (TR)] 94-01, Revision 2" of the NRC safety evaluation report in NEI 94-01, Revision 2-A, dated October 2008 (ADAMS Accession No. ML100620847), as the implementing document for the performance-based Option B of 10 CFR Part 50, Appendix J. These amendments allow CNP Units 1 and 2 to extend the Type A containment integrated leakage rate testing interval up to 15 years. The amendments also extend the Type C local leakage rate test interval from 60 months to 75 months.

A copy of our related safety evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Docket Nos. 50-315 and 50-316

Enclosures:

1. Amendment No. 326 to DPR-58

2. Amendment No. 309 to DPR-74

3. Safety Evaluation cc w/encls: Distribution via ListServ Sincerely, Allison W. Dietrich, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-315 DONALD C. COOK NUCLj:AR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 326 License No. DPR-58

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Indiana Michigan Power Company (the licensee) dated March 7, 2014, as supplemented by letters dated September 30, 2014, December 16, 2014, January 15, 2015, and February 20, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 1 O CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-58 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 326. are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days.

FO~/;;EGULATORY COMMISSION David L. Pelton, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-58 and Technical Specifications Date of Issuance: March 30, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 326 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 DOCKET NO. 50-315 Replace the following page of the Renewed Facility Operating License No. DPR-58 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

REMOVE INSERT 3

3 Replace the following page of Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

REMOVE INSERT 5.5-14 5.5-14 and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 1 O CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not to exceed 3304 megawatts thermal in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 326, are hereby incorporated in this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Less than Four Loop Operation The licensee shall not operate the reactor at power levels above P-7 (as defined in Table 3.3.1-1 of Specification 3.3.1 of Appendix A to this renewed operating license) with less than four reactor coolant loops in operation until (a) safety analyses for less than four loop operation have been submitted, and (b) approval for less than four loop operation at power levels above P-7 has been granted by the Commission by amendment of this license.

(4)

Fire Protection Program Indiana Michigan Power Company shall implement and maintain in effect all provisions of the approved fire protection program that comply with 1 O CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee's amendment request dated July 1, 2011, as supplemented by letters dated September 2, 2011, April 27, 2012, June 29, 2012, August 9, 2012, October 15, 2012, November 9, 2012, January 14, 2013, February 1, 2013, Renewed License No. DPR-58 Amendment No.-~. ~. 326

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 1 O CFR 50.54(0) and 1 O 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.

b.

The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 12 psig.

c.

The maximum allowable containment leakage rate, La, at Pa, shall be 0.25%

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 ares 0.60 La for the Type Band C tests and s 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.

Cook Nuclear Plant Unit 1 5.5-14 Amendment No. 2-8-7, 2-98, 326

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-316 DONALD C COOK NUCLEAR PLANT. UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 309 License No. DPR-7 4

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Indiana Michigan Power Company (the licensee) dated March 7, 2014, as supplemented by letters dated September 30, 2014, December 16, 2014, January 15, 2015, and February 20, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 1 O CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-74 is hereby amended to read as follows:

(2)

Tecnnical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 309, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days.

David L. Pelton, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-74 and Technical Specifications Date of Issuance: March 30, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 309 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-74 DOCKET NO. 50-316 Replace the following page of the Renewed Facility Operating License No. DPR-74 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

REMOVE INSERT 3

3 Replace the following page of Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

REMOVE INSERT 5.5-14 5.5-14 radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Pursuant to the Act and 1 O CFR Parts 30, 40, and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 1 O CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 1 O CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not to exceed 3468 megawatts thermal in accordance with the conditions specified herein and in Attachment 1 to the renewed operating license.

The preoperational tests, startup tests and other items identified in Attachment 1 to this renewed operating license shall be completed. Attachment 1 is an integral part of this renewed operating license.

(2) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 309, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Additional Conditions (a) Deleted by Amendment No. 76 (b) Deleted by Amendment No. 2 (c) Leak Testing of Emergency Core Cooling System Valves Indiana Michigan Power Company shall prior to completion of the first inservice testing interval leak test each of the two valves in series in the Renewed License No. DPR-7 4 Amendment No., -300, W-7, 309

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 5.5.15 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 1 O 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 1 O 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.

b.

The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa. is 12 psig.

c.

The maximum allowable containment leakage rate, La, at Pa, shall be 0.25%

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.

Battery Monitoring and Maintenance Program This program provides for battery restoration and maintenance, based on the recommendations of IEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," 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. 2W, 2-79, 309

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 326 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 AND AMENDMENT NO. 309 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-315 AND 50-316

1.0 INTRODUCTION

By application dated March 7, 2014 (Reference 1), as supplemented by letters dated September 30, 2014, December 16, 2014, January 15, 2015, and February 20, 2015, (References 2, 3, 4, and 5, respectively), Indiana Michigan Power Company (the licensee) requested a license amendment for Donald C. Cook Nuclear Plant (CNP), Units 1 and 2.

Specifically, the licensee requested to adopt Nuclear Energy Institute (NEI) 94-01, Revision 3-A (Reference 6), as the implementation document to develop the performance-based primary containment leakage testing program at CNP Units 1 and 2, in accordance with Title 1 O of the Code of Federal Regulations (10 CFR) Part 50, Appendix J, Option B. The proposed amendment would revise the CNP Units 1 and 2 technical specification (TS) 5.5.14, "Containment Leakage Rate Testing Program," by replacing the reference to Regulatory Guide (RG) 1.163 (September 1995) with a reference to NEI 94-01, Revision 3-A and Section 4.1, "Limitations and Conditions for NEI [Topical Report (TR)] 94-01, Revision 2" of the U.S. Nuclear Regulatory Commission (NRC) safety evaluation report (SER) in NEI 94-01, Revision 2-A, (Reference 7) dated October 2008, as the implementation document to develop the 1 O CFR Part 50, Appendix J, Option B. performance-based primary containment leakage testing program for CNP Units 1 and 2.

This amendment would allow CNP to extend its performance-based primary containment integrated leakage rate test (ILRT) or Type A test interval to no longer than 15 years.

Accordingly, the licensee has requested the current performance-based Type A test interval at CNP to be extended from 10 years to 15 years so that the next Type A test for CNP Units 1 and 2 can be conducted by November 1, 2021 (Unit 1) and April 22, 2021 (Unit 2), instead of the current due dates of November 1, 2016 (Unit 1), and April 22, 2016 (Unit 2). This amendment also requests to extend the Type C local leakage rate test interval from 60 months to 75 months.

The supplemental letters dated September 30, 2014, December 16, 2014, January 15, 2015, and February 20, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on May 27, 2014 (79 FR 30188)

2.0 REGULATORY EVALUATION

The regulations in 10 CFR 50.54(0) require that the primary reactor containments for water cooled power reactors shall be subject to the requirements set forth in Appendix J to 10 CFR Part 50, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors."

Appendix J to 10 CFR Part 50 includes two options, Option A - Prescriptive Requirements, and Option B - Performance-Based Requirements, either of which can be chosen for meeting the requirements of Appendix J. The testing requirements in Appendix J ensure that leakage through the primary reactor containment and related systems and components penetrating primary containment does not exceed allowable leakage rate value specified in the TSs or associated bases and that integrity of the containment structure is maintained during its service life.

The licensee has adopted and has been implementing Option B for meeting the requirements of Appendix J. Option B of Appendix J specifies the performance-based requirements and criteria for preoperational and subsequent leakage-rate testing. These requirements are met by performance of Type A tests to measure the containment system overall integrated leakage rate; Type B pneumatic tests to detect and measure local leakage rates across pressure retaining, leakage-limiting boundaries such as penetrations; and Type C pneumatic tests to measure containment isolation valve leakage rates. After the preoperational tests, these tests are required to be conducted at periodic intervals based on the historical performance of the overall containment system (for Type A tests), and based on the safety significance and historical performance of each boundary and isolation valve (for Type Band C tests) to ensure the integrity of the overall containment system as a barrier to fission product release. The leakage rate test results must not exceed the allowable leakage rate with margin, as specified in the technical specifications (TSs). Option B also requires that a general visual inspection for structural deterioration of the accessible interior and exterior surfaces of the containment, which may affect the containment leak-tight integrity, be conducted prior to each Type A test and at a periodic interval between tests based on the performance of the containment system.

Section V.B.3 of 10 CFR Part 50, Appendix J, Option B, requires that the regulatory guide or other implementation document used by a licensee to develop a performance-based leakage-testing program be included, by general reference, in the plant technical specifications.

Furthermore, the submittal for TS revisions must contain justification, including supporting analyses, if the licensee chooses to deviate from methods approved by the Commission and endorsed in a regulatory guide.

The implementation document that is currently referenced in the CNP TS 5.5.14, "Containment Leakage Rate Testing Program," is RG 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995. RG 1.163 endorsed NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 26, 1995, as a document that provides methods acceptable to the NRC staff for complying with the provisions of Option B of 10 CFR Part 50, Appendix J, subject to four regulatory positions delineated in Section C of the RG. NEI 94-01, Revision 0, includes provisions that allow the performance-based Type A test interval to be extended to 10 years, based upon two consecutive successful tests. The most recent two Type A tests at D.C. Cook have been successful, so the current interval requirement is 10 years. NEI 94-01, Revision 2-A, describes an approach for implementing the optional performance-based requirements of Option B of 10 CFR Part 50, Appendix J. It incorporates the regulatory positions stated in RG 1.163 (September 1995), and includes provisions for extending Type A test intervals to up to 15 years. In the NRC SER, dated June 25, 2008 (Reference 8), the NRC staff concluded that NEI TR 94-01, Revision 2, describes an acceptable approach for implementing the optional performance-based requirements of Option B of 10 CFR Part 50, Appendix J, and is acceptable for referencing by licensees proposing to amend their TS in regards to containment leakage rate testing, subject to the specific limitations and conditions listed in Section 4.1 of the SER The licensee also proposes to extend testing of Type C components from a maximum interval of 60 months to 75 months. Guidance for extending Type C Local Leakage Rate Test (LLRT) intervals beyond 60 months is given in NEI 94-01, Revision 3-A (Reference 6). NEI 94-01, Revision 3-A (page iv, Executive Summary) states that:

Intervals may be increased from 30 months up to a maximum of 120 months for Type B tests (except for containment airlocks) and up to a maximum of 75 months for Type C tests. If the Type Band C test results are not acceptable, the test frequency should be set at the initial test intervals. Once the cause determination and corrective actions have been completed, acceptable performance may be reestablished and the testing frequency returned to the extended intervals as specified in this document.

As described in NRC letter to NEI, "Request Revision to Topical Report NEI 94-01, Revision 3-A, 'Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J,"' dated August 20, 2013, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13192A394), Revision 3-A inadvertently did not include the six limitations and conditions provided in NRC's June 25, 2008, SER (Reference 8) approving NEI TR 94-01, Revision 2. The letter states that "Any licensee submissions referencing the TR

[Revision 3-A] will require requests for additional information from the NRC to address the limitations and conditions from the NRC [SER] for NEI 94-01, Revision 2." The licensee has addressed these limitations and conditions as discussed further in Section 3.0, below.

10 CFR 50.55a, "Codes and Standards," contains the containment inservice inspection (CISI) requirements that, in conjunction with the requirements of Appendix J, ensure the continued leak tightness and structural integrity of the containment during its service life.

1 O CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," states, in part, that the licensee shall monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components are capable of fulfilling their intended functions.

10 CFR 50.36(c)(3), "Surveillance requirements," states that surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

10 CFR 50.36(c)(5), "Administrative controls," states that administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in 10 CFR 50.4.

3.0 TECHNICAL EVALUATION

3.1 Proposed Changes CNP Unit 1 TS, Section 5.5.14, "Containment Leakage Rate Testing Program," currently states the following:

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(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 Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:

1.

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

2.

A one-time exception to the requirement to perform post-modification Type A testing is allowed for the steam generators and associated piping, as components of the containment barrier. For this case, ASME [American Society of Mechanical Engineers] Section XI leak testing will be used to verify the leak tightness of the repaired or modified portions of the containment barrier. Entry into MODES 3 and 4 following the extended outage that commenced in 1997 may be made to perform this testing.

CNP Unit 2 TS, Section 5.5.14, "Containment Leakage Rate Testing Program," currently states the following:

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(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 Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:

1.

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

In letters dated March 7, 2014 (Reference 1), and September 30, 2014 (Reference 2), the licensee proposed to delete subparagraphs "1" and "2" from Section 5.5.14 of CNP Unit 1 TS and subparagraph "1" from Section 5. 5.14 of CNP Unit 2 TS, and to revise the CNP Units 1 and 2 TS, Section 5.5.14, paragraph "a", as follows:

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 licensee justified the proposed changes by demonstrating the adequate performance of the CNP Units 1 and 2 containment based on historical plant-specific containment leakage testing program results and GISI program results and supported by a plant-specific risk assessment, consistent with the guidance in NEI 94-01, Revision 3-A, and the conditions and limitations contained in NEI 94-01, Revision 2-A.

3.2 Deterministic Considerations: Structural and Leak-Tight Integrity of the Containment As required by 10 CFR 50.54(0), the CNP containment is subject to the requirements set forth in 10 CFR Part 50, Appendix J. Option B of Appendix J requires that test intervals for Type A, Type B, and Type C testing be determined by using a performance-based approach. The existing CNP TS 5.5.14, "Containment Leakage Rate Testing Program," identifies RG 1.163, which endorses NEI 94-01, Revision 0, as the guidance for this program. The LAR proposes to change CNP TS 5.5.14 to identify NEI 94-01, Revision 3-A and Section 4.1, "Limitations and Conditions for NEI TR 94-01, Revision 2" of the NRC SER in NEI 94-01, Revision 2-A, dated October 2008 as the guidance for the program.

Revision 2-A of NEI 94-01 was issued in 2008, and included provisions for extending the ILRT Type A interval to 15 years subject to the limitations and conditions provided in the SER for Revision 2. Revision 3-A was issued in July 2012, and included guidance for extending the Type C LLRT interval to 75 months. Type A testing measures the overall leakage rate of the primary reactor containment. Type C testing ensures that individual containment isolation valves are essentially leak tight by assessing leakage potential and supporting decisions on appropriate maintenance. Type C test results are summed with the typically lesser contributions from the Type B tests each refueling outage and that combined total is assessed with a corresponding performance criterion.

The staff finds the conditions and limitations contained in the staff's SER incorporated into NEI 94-01, Revision 2-A as well as the NEI 94-01, Rev 3-A, to be an acceptable complete reference for program guidance by licensees proposing to amend their TSs to permanently extend the ILRT surveillance (Type A) interval to 15 years and their Type C test interval to 75 months.

3.2.1 Adoption of NEI 94-01, Revision 3-A The NRC staff evaluated whether the licensee's application, as supplemented, adequately addressed and satisfied the limitations and conditions described in Section 4.1, "Limitations and Conditions for NEI TR 94-01, Revision 2" of the NRC SER in NEI 94-01, Revision 2-A (Reference 7).

a. NRC Condition 1 states:

For calculating the Type A leakage rate, the licensee should use the definition in the NEI TR 94-01, Revision 2, in lieu of that in ANSl/ANS-56.8-2002.

The licensee stated in Section 4.0 of the LAR (Reference 1) that:

following the NRC approval of this LAR, CNP will use the definition in Section 5.0 of NEI 94-01, Revision 3-A, for calculating the Type A leakage rate when future CNP Type A tests are performed.

NEI 94-01, Revision 3-A contains the same definition as in Revision 2-A for calculating the Type A test leakage rate. Therefore, the NRC staff finds that the licensee has adequately addressed Condition 1 of the NRC staff SER for NEI TR 94-01, Revision 2.

b. NRC Condition 2 states:

The licensee submits a schedule of containment inspections to be performed prior to and between Type A tests.

NEI 94-01, Revision 2-A, Section 9.2.3.2, "Supplemental Inspection Requirements," states that in order to provide continuing supplemental means of identifying potential containment degradation, a general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leak-tight integrity must be conducted prior to each Type A test and during at least three other outages before the next Type A test if the interval for the Type A test is extended to 15 years.

The licensee stated that the examinations performed in accordance with the CNP ASME Boiler and Pressure Vessel Code (Code),Section XI, Subsection IWE/IWL program satisfy the general visual examination requirements specified in 10 CFR Part 50, Appendix J, Option B.

In Section 4.4 of the LAR (Reference 1 ), the licensee stated that the frequency of examinations per Subsection IWE (three examinations over a 10-year interval) assures that at least three general visual examinations of metallic components will be conducted between the Type A tests, if the Type A test interval is extended to 15 years.

Furthermore, the licensee stated that (1) visual examinations of accessible concrete surfaces performed in accordance with Subsection IWL at a frequency of 5 years will result in at least two IWL examinations being performed during a 15-year Type A test interval; and (2) in addition to the IWE and IWL examinations, the licensee performs a visual inspection of the accessible interior and exterior concrete surfaces of the CNP containment structure prior to any Type A test. This examination is performed, in accordance with approved plant procedures to satisfy the requirements of the 10 CFR Part 50, Appendix J, testing program, and in sufficient detail to identify any evidence of deterioration which may affect the structural integrity or leak tightness of the containment building.

Additionally, the licensee provided the dates of completed and scheduled ILRTs, completed containment surface examinations, and an approximate schedule for future general visual examinations of containment surfaces, representative of a typical 15-year period between Type A tests for both CNP Units 1 and 2.

The licensee concluded in Section 4.4 of the LAR that:

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

On the basis that the licensee's schedule of general visual examinations, described in the LAR, results in at least three examinations between Type A tests and one examination immediately prior to the Type A test for both containment concrete and metallic liner surfaces, the NRC staff concludes that the licensee's inspection schedule plan meets the general visual examination requirements in Section 9.2.3.2 of NEI 94-01, Revision 2-A, and 10 CFR Part 50, Appendix J, Option B, and therefore, satisfies Condition 2 in the NRC staff SER for NEI TR 94-01, Revision 2.

c.

NRC Condition 3 states:

The licensee addresses the areas of the containment structure potentially subjected to degradation.

In Section 4.4.1, "IWE Examinations" and Section 4.4.2 "IWL Examinations" of the LAR, and letter dated September 30, 2014 (Reference 2), the licensee stated that (1) general visual examinations of accessible interior and exterior surfaces of the containment are conducted in accordance with the CNP CISI program and schedule, which implements the requirements of the ASME Code,Section XI, Subsections IWE and IWL, as required by 10 CFR 50.55a(g); (2) the CNP Units 1 and 2 containments do employ moisture barriers but do not employ bellows on penetrations through the containment pressure retaining boundaries; and (3) the CNP CISI program contains requirements to evaluate the acceptability of the inaccessible areas if such conditions were identified, in accordance with 1 O CFR 50.55a(b)(2)(ix)(A) and 1 O CFR

50. 55a(b)(2)(viii)(E).

In its LAR (Reference 1), the licensee provided the following information relative to the first 10-year interval IWE examination:

No areas were deemed susceptible to accelerated degradation and aging; therefore, augmented examinations per Category E-C were not required.

In 1998, a visual examination identified corrosion and pitting of the steel liner plate along the moisture barrier seal near the interface of the concrete base floor and the containment wall liner plate in both CNP Units 1 and 2. Based on a detailed engineering analysis, it was concluded that the structural integrity and the leak tight integrity of the containments was not affected by the as-found condition of the liner.

The corrective action for the liner corrosion included modifying the moisture barrier seal design to prevent moisture intrusion and re-coating the affected area. Sections of the redesigned moisture barrier seal were removed in both Unit 1 and Unit 2 approximately 3 years after installation, and a visual examination was performed on the liner area where corrosion was previously identified. The visual examination found no moisture intrusion and no active corrosion. Continued monitoring of the moisture barrier seal is performed with the scheduled VT-3 visual examination of the moisture barrier seal area each IWE inspection period.

The NRC staff notes that the licensee root cause evaluation and corrective actions taken for the containment liner plate corrosion and pitting identified in both CNP Units 1 and 2 have been reviewed and discussed in NRC Inspection Reports 50-315/99026 and 50-316/99026 (ADAMS Accession No. ML003677533), and found to be acceptable.

In 1999, a visual examination of the containment liner identified an apparent weld repair of the liner plate. Surface preparation to allow further inspection dislodged repair material exposing a hole through the liner plate. It appeared that the liner hole resulted from an inadequate repair of a hole drilled in error during plant construction. After the damaged liner plate section was cut out, a piece of wood, determined to be the handle of a wire brush, was found embedded in the concrete.

Some minor corrosion was noted on the concrete side of the liner plate in the area of the embedded wire brush. The through-wall hole in the liner plate did not invoke the requirements for augmented examination per IWE-1240 since the hole was not the result of degradation or aging. The corrective action for the through-wall hole included removal of the wire brush and wooden handle to the maximum extent practical, performing a repair of the concrete, and replacing the liner section that contained the hole. The repair was vacuum box tested and subjected to an LLRT.

Furthermore, in Section 4.4.1, "IWE Examinations" of the LAR, the licensee stated that the code of record of CNP Units 1 and 2 for the second 10-year interval IWE examination is the 2004 Edition of the ASME Code,Section XI as modified by the 10 CFR 50.55a(b) limitations. In the second 10-year interval IWE examination, which commenced March 1, 2010, the licensee identified a glycol pipe penetration in Unit 1 with a large amount of wet discoloration due to condensation below the insulated piping. After insulation removal, the licensee performed a VT-1 visual examination which revealed minor pitting that did not impact the leak tightness or structural integrity of the containment boundary. The licensee classified this area as an augmented examination per Category E-C for continued monitoring with a VT-1 visual examination during successive inspection periods.

In its letter dated December 16, 2014 (Reference 3), the licensee responded to a staff request for additional information (RAI), stating that during the fall 2014 U 1 C26 refueling outage, an augmented VT-1 visual examination was performed on Unit 1 glycol pipe penetration 1-CPN-56.

The VT-1 examination revealed no change from the previous examination performed in the 2011 U1 C24 refueling outage. There was minor pitting on the west side of the penetration with less than 1 /32 inches material loss and minor pitting on the non-safety related weld pressurization channel with less than 1 /16 inches material loss. The licensee evaluated the examination results and concluded that the leak tightness and structural integrity of the Unit 1 containment structure was not adversely affected.

The licensee stated, in the LAR, that other than the augmented examination Category E-C discussed above, the remaining examinations are based on Category E-A and are visual (General, VT-3, and VT-1) examinations based on the ASME Code or 10 CFR rule requirements.

Based on the above information regarding the IWE and IWL examinations of the CNP Units 1 and 2 containment structures and the CNP operating experience, to date, no conditions that would indicate the presence of any significant degraded condition in the accessible or inaccessible areas of the containment structure and steel liner affecting the leak tightness or structural integrity of the CNP Units 1 and 2 containment structures have been identified. In addition, the CNP GISI program contains requirements to evaluate the acceptability of the inaccessible areas if such conditions were identified, in accordance with 10 CFR 50.55a(b)(2)(ix)(A) and 10 CFR 50.55a(b)(2)(viii)(E). As such, the NRC staff concludes that the licensee has adequately addressed the intent of Condition 3 of the NRC staff SER for NEI TR 94-01, Revision 2.

d. NRC Condition 4 states:

The licensee addresses any tests and inspections performed following major modifications to the containment structure, as applicable.

The licensee stated in the LAR (Reference 1, Section 4.0), that "CNP has already replaced the steam generators and the reactor vessel closure heads."

Furthermore, the licensee stated in the LAR (Reference 1, Section 4.2) that (1) no modifications that require a Type A test are planned prior to Unit 1 U1 C30 (fall 2020) and Unit 2 U2C26 (spring 2021) refueling outages, when the next Type A tests would be performed in accordance with this proposed change; (2) any unplanned modifications to the containment building prior to the next scheduled Type A test would be subject to the special testing requirements of Section IV.A of 1 O CFR Part 50, Appendix J; (3) there have been no pressure or temperature excursions in either Unit 1 or Unit 2 containments which could have adversely affected reactor building integrity; and (4) there is no anticipated addition or removal of plant hardware within either Unit 1 or Unit 2 containment which could affect its leak-tightness.

Based on the above, the NRC staff concludes that the licensee's program will implement the staff position with regard to post-repair pressure testing following major and minor containment repairs and modifications, as explained in Section 3.1.4 of the staff SER for NEI TR 94-01, Revision 2. Therefore, the NRC staff concludes that the licensee has adequately addressed Condition 4 of the NRC staff SER for NEI TR 94-01, Revision 2.

e. NRC Condition 5 states:

The normal Type A test interval should be less than 15 years. If a licensee has to utilize the provision of Section 9.1 of NEI TR 94-01, Revision 2, related to extending the ILRT interval beyond 15 years, the licensee must demonstrate to the NRC staff that it is an unforeseen emergent condition.

The licensee stated, in Section 4 of the LAR, that it acknowledges and accepts the NRC staff position in Condition 5, as communicated to the nuclear industry in NRC 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," dated December 8, 2008 (ADAMS Accession No. ML080020394).

Accordingly, the NRC staff concludes that the licensee has confirmed its understanding that any extension of the Type A test interval beyond the performance-based limit of 15 years should be infrequent and should be requested only for compelling reasons, and that the staff will implement the position in RIS 2008-27 in reviewing such LARs. Therefore, the NRC staff concludes that the licensee has adequately addressed Condition 5 of the staff SER for NEI TR 94-01, Revision 2.

f.

NRC Condition 6 states:

For plants licensed under 10 CFR Part 52, applications requesting a permanent extension of the ILRT surveillance interval to 15 years should be deferred until after the construction and testing of containments for that design have been completed and applicants have confirmed the applicability of NEI TR 94-01, Revision 2, and EPRI Report No.1009325, Revision 2, including the use of past containment ILRT data.

The licensee stated in its LAR that this condition is not applicable to CNP since CNP is not licensed under 10 CFR Part 52. The NRC staff concludes that Condition 6 does not apply to CNP since CNP is currently an operating reactor licensed under 10 CFR Part 50.

3.2.1.1 Conclusion for Licensee's Adoption of NEI 94-01, Revision 3-A Based on the above evaluation, the NRC staff conCludes that the licensee has adequately addressed and satisfied the six conditions in Section 4.1 of the NRC staff SER for NEI TR 94-01, Revision 2. Therefore, the staff finds it acceptable for CNP to adopt NEI 94-01, Revision 3-A, as the implementation document in its TS 5.5.14, "Containment Leakage Rate Testing Program."

3.2.2 Extension of Type A Test Interval from 10 to 15 Years 3.2.2.1 Description of CNP Primary Containment System As described in Section 5.0 of the CNP updated safety analysis report, CNP Units 1 and 2 utilize a pressurized-water reactor with ice condenser-type containment. The containment building is a reinforced concrete structure consisting of a vertical cylinder, a hemispherical dome, and a flat base slab. A welded steel liner is attached to the inside face of the concrete (shell, dome, and the base slab) to ensure a high degree of leak tightness. The foundation slab liner plate is installed on the top face of the slab and is covered with a concrete mat. Each containment structure is penetrated by access penetrations, process piping and electrical penetrations.

The CNP leak-tight integrity of the penetrations and isolation valves are verified through LLRT (Type Band Type C tests) and the overall leak-tight integrity and structural integrity of the primary containment is verified through an ILRT (Type A test), as required by 10 CFR Part 50, Appendix J. The leakage rate testing requirements of 10 CFR Part 50, Appendix J, Option B (Type A, Type B, and Type C Tests) and the CISI requirements mandated by 10 CFR 50.55a, together, ensure the continued leak-tight and structural integrity of the containment during its service life.

3.2.2.2 Historical Plant-Specific Containment Leakage Testing Program Results As indicated in the LAR (Reference 1) and the CNP TS 5. 5.14, the maximum allowable containment leakage rate, La, is 0.25 percent of containment air weight per day at the peak calculated containment internal pressure for design basis loss-of-coolant accident, Pa.

In Section 4.2 of the LAR, the licensee provided the as-found leakage of the last three CNP Units 1 and 2 Type A test results performed in 1989, 1992, and 2006. The maximum as-found leakage was 42 percent of the acceptance limit of 1.0 La. The 2006 as-found leakage was approximately 34 percent and 24 percent of the acceptance limit of 1.0 La for Unit 1 and Unit 2, respectively.

Furthermore, the licensee stated that the CNP Appendix J, Type B and Type C testing program requires testing of electrical penetrations, air locks, hatches, flanges, and valves within the scope of the program as required by 10 CFR Part 50, Appendix J, Option B and CNP TS 5.5.14.

The licensee provided the combined Type B and Type C leakage rate for the last three refueling outages and stated that the current total penetration leakage on a minimum path basis is less than 20 percent of the leakage allowed for containment integrity. The licensee also indicated that the as-found combined Type B and Type C leakage rate, on a minimum pathway basis, for the Unit 2 U2C20 (April 2012) refueling outage exceeded the allowable leakage. Section 4.3 of the LAR indicates that the licensee performed an evaluation and appropriate repair was implemented to correct the leaking component.

The licensee stated that industry experience has shown that Type Band Type C tests can identify the vast majority (over 95 percent) of all potential primary containment leakage paths.

The licensee stated that this LAR adopts the guidance in NEI 94-01, Revision 3-A, in place of NEI 94-01, Revision 0, but otherwise does not affect the scope or performance of Type B or Type C tests, and that Type B and Type C testing will continue to provide a high degree of assurance that primary containment integrity is maintained.

In Section 4.3 of the LAR, the licensee indicated that (1) each unit of CNP has 120 mechanical penetrations and 55 electrical penetrations that are subject to performance-based Type B or Type C testing; and (2) currently, approximately 5 percent of the components (Unit 1 and Unit 2 combined) require testing at an increased frequency.

Based on the information discussed above, the NRG staff concludes that (1) the performance history of successful completion of the last three Type A tests supports extending the current ILRT interval to 15 years; (2) the current combined leakage from the Type Band Type C tests is below 20 percent of the acceptance limit; (3) only 5 percent of CNP Units 1 and 2 mechanical and electrical penetrations that are subject to performance-based Type B or Type C testing require testing at an increased frequency interval, which demonstrates good performance of Type Band Type C penetrations at CNP; (4) the licensee has appropriately taken corrective actions for those components that failed their allowable leak rate limit; and (5) there is reasonable assurance that the licensee is effectively implementing its Type B and Type C testing program under Option B of Appendix J to 10 CFR Part 50. These conclusions support approving the extension of the Type A test interval at CNP from 1 O to 15 years. Thus, the next Type A testing interval at CNP may be conducted no later than November 1, 2021 (Unit 1) and April 22, 2021 (Unit 2), in lieu of the current due date of November 1, 2016 (Unit 1) and April 22, 2016 (Unit 2).

3.2.3 Containment In-Service Inspection Program In Section 4.0 of the LAR, the licensee stated, "General visual examination of accessible interior and exterior surfaces of the containment system for structural problems is conducted in accordance with the CNP IWE/IWL [GISI] Plans which implement the requirements of the ASME Code,Section XI, Subsections IWE and IWL, ~s required by 10 CFR 50.55a(g)." The IWE/IWL inspections and supplemental inspections, in accordance with other approved plant procedures, are used to satisfy the general visual examination requirements of Appendix J, Option B and to monitor and manage the age-related degradations of the primary containment to ensure that containment structural and leak-tight integrity is maintained through its service life.

The operating experience relative to the moisture barrier at the interface of the concrete base floor and the containment wall liner plate has been discussed in Section 3.2.1 (c) of this safety evaluation and will not be repeated here.

In Section 4.4.1 of the LAR, the licensee stated that the inspection of the containment liner coating during the second 10-year interval IWE examination revealed only "some flaking and discoloration of coatings along with some surface corrosion and minor pitting which did not impact the leak tightness or structural integrity of the containment boundary."

In response to an NRG staff RAI regarding the applicability of the NRG Information Notice 2014-07, "Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner," the licensee indicated, in a letter dated September 30, 2014 (Reference 2), that (1) CNP Containment Penetration and Weld Channel Pressurization (CPWCP) system piping design and operating configuration does not make it vulnerable to allow moisture to reach the liner; (2) CNP CPWCP system lines are on manifolds and not below floor level with covers as presented in the three plant examples from NRC Information Notice 2014-07; (3) the CPWCP system piping is opened only during the ILRT to allow containment atmosphere to reach all areas of the liner; and (4) CPWCP system piping is capped and closed (normal configuration) between tests to prevent atmospheric moisture intrusion into the piping.

In Section 4.4.2 of the LAR, the licensee stated that, as part of the second 10-year CISI Program, CNP Units 1 and 2 have completed the IWL examination of the containment structures for Units 1 and 2 by the required date of August 24, 2011, in accordance with the requirements of the 2004 Edition of the ASME Code,Section XI. These examinations on the concrete exterior surfaces were conducted under the direction of the Responsible Engineer using the visual (VT-3C and VT-1C) method. The conditions noted during the 2011 IWL examination included cracking, efflorescence, popouts, spalling, exposed rebar, and embedded foreign material. The licensee stated that examination of the CNP Unit 1 and Unit 2 containment structures did not reveal any significant observations that can potentially affect the containment structural integrity. Code repairs of the concrete have been performed for conditions such as exposed rebar and popout areas. Other areas that were not repaired will continue to be monitored in future examinations.

Based on the results of the recent IWE/IWL inspections discussed above, the NRC staff finds that there has not been evidence to date of significant degradation of the CNP Units 1 and 2 containment structures, and that the degradations noted have been entered into the CNP corrective action program, and appropriately managed and/or corrected. Based on the above evaluation, the staff finds that there is reasonable assurance that the licensee is adequately implementing the CNP CISI program to monitor and manage age-related degradation of the CNP containment structure.

3.2.4 Extension of Type C Test Interval from 60 to 75 months The LAR provided the following historical combined LLRT (Type B and Type C) results.

Combined As-Found Minimum Pathway As-Found Minimum Pathway LLRT Test Leakage Rate in Standard Leakage Rate as a Percentage Completion Cubic Centimeters per Minute of 0.6 La Date (seem}

(0.6 La is the Performance (La is 110,119 seem)

Criterion for Evaluating Past Acceptability of Combined Penetration Leakage)

Unit 1 April 2010 41,616 62.9 September 2011 46, 107 69.7 May 2013 9,838 14.9 Unit 2 November 2010 35,845 54.2 April 2012

>La N/A November 2013 11, 193 16.9 The tests performed in 2010, 2011, and 2013 for Unit 1, and in 2010 and 2013 for Unit 2, showed leakage rates to be less than the maximum allowable containment leakage rate 0.6La with margin sufficient to suggest there is reasonable assurance that the proposed testing interval extension can be accommodated safely.

By supplemental "AEP-NRC-2015-25, Response to RAI Regarding License Amendment Request to Revise Technical Specification 5.5.14, Containment Leakage Rate Testing Program, to Use NEI-94-01, Revision 3-A as Regulatory Guidance Versus Current [[NEI" contains a listed "[" character as part of the property label and has therefore been classified as invalid., Rev. 0|letter dated February 20, 2015]] (Reference 5), the licensee provided the following additional information regarding the Unit 2 April 2012 Type Band C Test combined as-found minimum pathway leakage total.

The as-found combined Type B and Type C tests minimum pathway leakage rate for Unit 2 in April 2012 was 11,286 (about 17.1 percent of the 0.6 La performance criterion) if penetration 2-CPN-15 contribution was excluded. Penetration 2-CPN-15 accommodates the common discharge header for six emergency core cooling system relief valves allowing the routing of potentially highly contaminated fluids to the pressurizer relief tank inside the containment. The containment isolation barriers for this penetration consist of a closed loop outside containment and a check valve, 2-Sl-189 inside containment. The leakage through 2-Sl-189 was so large that the test apparatus could not achieve the required test pressure ( 12 pounds per square inch gage). The CNP leakage testing program directed that the leakage rate measured for 2-Sl-189 be recorded as both the minimum pathway leakage rate for penetration 2-CPN-15 for evaluation to the as-found performance criterion and as the maximum pathway leakage rate for evaluation with the acceptance criterion for re-establishing containment operability. CNP corrective action program Action Request 2012-5255 documented the determination of past containment operability on the basis that the closed loop outside containment penetration barrier had been maintained intact and would have prevented any containment atmospheric leakage out of containment and thus the performance criterion for as-found minimum pathway combined Type B and Type C total was actually met although not recorded as such by the test program procedure. Industry guidance document ANS-56.2 I ANSI N271-1976, "Containment Isolation Provisions for Fluid Systems" provides a definition and requirements for closed systems used as a penetration barrier in lieu of an isolation valve. ANSI N271, Section 3.6.7 identifies a number of design elements intended to ensure that a closed system is sufficiently robust for use as a penetration barrier. ANSI N271, Section 3.6.4 contains an additional provision that the closed system be Appendix J tested unless it can be shown by inspection that system integrity is being maintained with the system operating at a pressure equal to or greater than containment design pressure. This latter provision is usually accomplished by observing no external leakage with the system in operation or otherwise sufficiently pressurized. The LAR stated that the leaking check valve 2-Sl-189 was repaired by implementing a (vendor recommended) modification to install longer hinge ring retaining dowel pins (to ensure better and more reliable disk seating).

Based on the historical LLRT test results and the ample margins to the performance criteria in conjunction with the containment inspection programs, the staff finds that, consistent with the guidance in NEI 94-01, Revision 3-A, and the limitations and conditions from section 4.1 of the staff's SE for NEI 94-01, Revision 2, there is reasonable assurance that the extension of Type C testing interval to 75 months would not result in a challenge to the performance criteria, and thus this interval change is acceptable.

3.3 Probabilistic Risk Assessment 3.3.1

Background

Section 9.2.3.1, "General Requirements for ILRT Interval Extensions beyond Ten Years," of NEI 94-01, Revision 3-A (Reference 6), states that plant-specific confirmatory analyses are required when extending the Type A ILRT interval beyond ten years. Section 9.2.3.4, "Plant-Specific Confirmatory Analyses," of NEI 94-01, Revision 3-A, states that the assessment should be performed using the approach and methodology described in Electric Power Research Institute (EPRI) Technical Report No. 1018243, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals" (Reference 9). The analysis is to be performed by the licensee and retained in the plant documentation and records as part of the basis for extending the ILRT interval.

In its SER, dated June 25, 2008 (Reference 8), the NRC staff found the methodology in NEI TR 94-01, Revision 2, and EPRI Technical Report No. 1009325, Revision 2, acceptable for referencing by licensees proposing to amend their TS to permanently extend the ILRT interval to 15 years, provided certain conditions are satisfied. These conditions, set forth in Section 4.2 of the SER for EPRI Technical Report No. 1009325, Revision 2, stipulate that:

1. The licensee submit documentation indicating that the technical adequacy of their Probabilistic Risk Assessment (PRA) is consistent with the requirements of Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," relevant to the ILRT extension application. Additional application specific guidance on the technical adequacy of a PRA used to extend ILRT intervals is provided in the SER for EPRI Technical Report No. 1009325, Revision 2.

2. The licensee submit documentation indicating that the estimated risk increase associated with permanently extending the I LRT surveillance interval to 15 years is small and consistent with the clarification provided in Section 3.2.4.5 of the SER for EPRI Technical Report No. 1009325, Revision 2.

3. The methodology in EPRI Technical Report No. 1009325, Revision 2, is acceptable provided that the average leak rate for the pre-existing containment large leak accident case (i.e., accident case 3b) used by licensees is assigned a value of 100 times the maximum allowable leakage rate (La) instead of 35 La.

4. A LAR is required in instances where containment over-pressure is relied upon for emergency core cooling (ECCS) system performance.

3.3.2 Plant-Specific Risk Evaluation The licensee performed a risk impact assessment for extending the Type A containment ILRT interval to once in 15 years. The risk assessment was provided in Enclosure 3 of the LAR (Reference 1 ). Additional information was provided by the licensee in response to NRC RAls in letters dated January 15, 2015 (Reference 4), and February 20, 2015 (Reference 5).

In Section 4.6.1 of Enclosure 2 of the LAR, the licensee stated that the plant-specific risk assessment follows the guidance in NEI 94-01, Revision 3-A (Reference 6); the methodology described in EPRI Technical Report No. 1018243, Revision 2-A of Technical Report No 1009325; and the NRC regulatory guidance outlined in RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." Additionally, the licensee used the methodology from Calvert Cliffs Nuclear Power Plant to assess the risk from undetected containment leaks due to steel liner corrosion.

The licensee addressed each of the four conditions for the use of EPRI Technical Report No. 1009325, Revision 2, which are listed in Section 4.2 of the NRC SER for NEI TR 94-01, Revision 2. A summary of how each condition has been met is provided in the sections below.

3.3.2.1 Technical Adequacy of the PRA The first condition stipulates that the licensee submit documentation indicating that the technical adequacy of their PRA is consistent with the requirements of RG 1.200 relevant to the ILRT extension application.

Consistent with the information provided in RIS 2007-06 (ADAMS Accession No. ML070650428), "Regulatory Guide 1.200 Implementation," the NRC staff will use Revision 2 of RG 1.200 (ADAMS Accession No. ML090410014) to assess the technical adequacy of the PRA used to support risk-informed applications received after March 2010. In Section 3.2.4.1 of the SER for NEI TR 94-01, Revision 2 and EPRI Technical Report No. 1009325, Revision 2, the NRC staff states that Capability Category I of the ASME PRA standard shall be applied as the standard for assessing PRA quality for IRL T extension applications, since approximate values of core damage frequency (CDF) and large early release frequency (LERF) and their distribution among release categories are sufficient to support the evaluation of changes to ILRT frequencies.

Section 4.6.2 of Enclosure 2 to the LAR states that an updated internal events PRA model, designated by the licensee as "09MORW + T&M corrections," and an updated Level 2 model were used to obtain the risk metrics associated with the ILRT extension. These models were updated to address the findings from the 2012 gap assessment. The licensee states in to the LAR that the PRA model is maintained and updated under a PRA configuration control program in accordance with CNP procedures. The LAR further states that plant changes are reviewed and the PRA model is periodically updated to reflect such changes.

As indicated in Enclosure 6 to the LAR, the internal events PRA model was peer reviewed in 2001 and 2009 by the Westinghouse Owners Group. Enclosure 6 further states that a gap assessment was conducted in 2012, which was performed against RG 1.200, Revision 2. The licensee additionally states that a focused-scope peer review of the updated Level 2 model was performed in 2013.

The licensee provided the status of the facts and observations (F&Os) from the early peer reviews of 2001 and 2009 in Table 1.0 of Enclosure 6 to the LAR. In response to an RAI dated January 15, 2015 (Reference 4), the licensee performed a review of the findings from the 2012 gap assessment and the 2013 focused-scope peer review to identify supporting requirements (SRs) of the PRA standard not meeting Capability Category I. As described in the RAI response dated January 15, 2015, and clarified in the follow up RAI response dated February 20, 2015 (Reference 5), the licensee performed this review against the 2009 PRA standard ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2, but justified the impact on the ILRT extension application against the comparable requirements of the 2003 PRA standard ASME RA-Sa-2003. The licensee identified and explained the resolution or the impact of eleven SRs that did not meet Capability Category I requirements of the ASME PRA Standard.

Two of the eleven SRs (SRs LE-A4 and LE-05 from the 2003 revision of the ASME Standard),

identified by the 2012 gap assessment, were subsequently closed and verified by the 2013 focused-scope peer review. One SR (SR LE-G6 from the 2009 revision of the ASME Standard) was identified to meet Capability Category I. The remaining eight of eleven SRs were assessed by the licensee to be related to documentation (SRs IE-C10, AS-C3, QU-E2, and LE-G8 from the 2003 revision of the ASME Standard) or increased level of conservatism (SRs SC-C2, SC-C3, LE-F2, and LE-G7 from the 2003 revision of the ASME Standard), and therefore as having no impact on the ILRT extension application. The NRC staff reviewed the disposition and assessment of identified SRs by the licensee and finds that the requirements of the 2003 and the 2009 revisions of the PRA standard are equivalent for those SRs. Therefore, the staff finds the licensee's assessment of those SRs acceptable for supporting this application.

In Section 6.6 of Enclosure 3 of the LAR, the licensee performed an analysis of the impact of external events, which included an evaluation of seismic and fire risks. The licensee stated that the Individual Plant Examination for External Events (IPEEE) found that only internal fires and seismic events were significant, and other external events such as high winds, external floods, transportation, and accidents at nearby facilities were considered and screened in the IPEEE.

As such, the licensee considered the risk associated with the latter external events as negligible.

Section 2.0 of Enclosure 3 of the LAR states that the licensee used the PRA results from the IPEEE to assess the seismic risk contribution. It further states that the seismic PRA model has not been updated since the IPEEE. To assess the fire risk contribution, the licensee used the results from the fire PRA submitted in support of its transition to the National Fire Protection Association (NFPA) 805 Standard. The licensee states that the fire CDF values used in the analysis "correspond to the post-transition plant, and do not include credit for low-leakage RCP seals." The licensee states that "[t)he use of the post-transition Internal Fire CDF values is considered reasonable since the transition to NFPA-805 must be completed within one year after the receipt of the Safety Evaluation from the NRC." The safety evaluation allowing transition to NFPA-805 was issued on October 24, 2013 (ADAMS Accession No. ML13140A398) and the one-year period has already been reached. Therefore, the staff finds the use of the NFPA-805 post-transition internal fire CDF values acceptable for assessing the risk of the ILRT extension.

In Section 3.2.4.2 of the SER for NEI TR 94-01, Revision 2 and EPRI Technical Report No. 1009325, Revision 2, the NRC staff states that:

Although the emphasis of the quantitative evaluation is on the risk impact from internal events, the guidance in EPRI Report No. 1009325, Revision 2, Section 4.2.7, "External Events," states that: "Where possible, the analysis should include a quantitative assessment of the contribution of external events (e.g., fire and seismic) in the risk impact assessment for extended ILRT intervals." This section also states that: "If the external event analysis is not of sufficient quality or detail to directly apply the methodology provided in this document [(i.e., EPRI Report No. 1009325, Revision 2)), the quality or detail will be increased or a suitable estimate of the risk impact from the external events should be performed." This assessment can be taken from existing, previously submitted and approved analyses or other alternate method of assessing an order of magnitude estimate for contribution of the external event to the impact of the changed interval.

The information used to estimate the effect on total LERF due to external events is considered acceptable because the increase in LERF was determined to meet the guidelines in RG 1.17 4, as discussed in Section 3.3.2.2 of this safety evaluation.

The licensee has evaluated its PRA against the current ASME PRA standard and Revision 2 of RG 1.200. It also addressed the findings from the peer reviews that are applicable to the ILRT interval extension. Therefore, the NRC staff concludes that the PRA model used by the licensee is of sufficient technical adequacy to support the evaluation of changes to ILRT frequencies. Accordingly, the first condition is met.

3.3.2.2 Estimated Risk Increase The second condition stipulates that the licensee submit documentation indicating that the estimated risk increase associated with permanently extending the I LRT interval to 15 years is small, and consistent with the guidance in RG 1.17 4 and the clarification provided in Section 3.2.4.5 of the NRC SER for NEI TR 94-01, Revision 2, and EPRI Technical Report No. 1009325, Revision 2. Specifically, a small increase in population dose should be defined as an increase in population dose of less than or equal to either 1.0 person-rem per year or 1 percent of the total population dose, whichever is less restrictive. In addition, a small increase in conditional containment failure probability (CCFP) should be defined as a value marginally greater than that accepted in previous one-time 15-year I LRT extension requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage points. Additionally, for plants that rely on containment over-pressure for net positive suction (NPSH) for ECCS injection, both CDF and LERF will be considered in the ILRT evaluation and compared with the risk acceptance guidelines in RG 1.17 4. As discussed in Section 3.2.4 of this SER, CNP Units 1 and 2 do not rely on containment over-pressure for ECCS performance.

Thus, the associated risk metrics include LERF, population dose, and CCFP.

The licensee provided the results of the plant-specific risk assessment in Section 4.6.3 of to the LAR. Details of the risk assessment are provided in Enclosure 3. The risk assessment is applicable to both CNP Unit 1 and Unit 2. The reported risk impacts are based on a change in test frequency from three tests in 10 years (the test frequency under 10 CFR Part 50 Appendix J, Option A) to one test in 15 years. The following conclusions can be drawn from the licensee's analysis associated with extending the Type A ILRT frequency:

1. The reported increase in LERF for internal events is 5.0BE-08/year for both units. These changes are considered to be "very small" (i.e., below 1 E-07 /year) per the acceptance guidelines in RG 1.17 4. The reported increase in LERF for internal and external events combined is 6.03E-07/year for Unit 1 and 5.62E-07/year for Unit 2. The risk contribution from external events includes the effects of internal fires and seismic events, as discussed in Section 3.3.2.1 of this Safety Evaluation. These changes in risk are considered to be "small" (i.e., between 1 E-06/year and 1 E-07/year) per the acceptance guidelines in RG 1.17 4. An assessment of total baseline LERF is also required to show that the total LERF is less than 1 E-05/year. The total LERF, including internal and external events, is estimated by the licensee to be 9.60E-06/year for both units, which is below 1 E-05 per year.

2. The requested change in Type A ILRT frequency from three in 10 years to once in 15 years results in an increase in the total population dose of 6. 92E-3 person-rem/year or 0.02 percent of the total population dose for both units. The reported increase in population dose is below the values provided in EPRI Technical Report No. 1009325, Revision 2-A, and defined in Section 3.2.4.6 of the NRC SER for NEI TR 94-01, Revision

2. Thus, this increase in the total integrated plant risk for the proposed change is considered small and supportive of the proposed change.

3. The increase in CCFP due to a change in test frequency from three in 1 O years to once in 15 years is 0. 32 percent for both units. This value is below the acceptance guidelines in Section 3.2.4.6 of the NRC SER for NEI TR 94-01, Revision 2.

Based on the risk assessment results, the NRC staff concludes that the increase in LERF is small and consistent with the acceptance guidelines of RG 1.17 4, and the increase in the total integrated plant risk and the magnitude of the change in the CCFP for the proposed change are small and supportive of the LAR. The defense-in-depth philosophy is maintained as the independence of barriers will not be degraded as a result of the requested change, and the use of the three quantitative risk metrics collectively ensures that the balance between prevention of core damage, prevention of containment failure, and consequence mitigation is preserved.

Accordingly, the second condition is met.

3.3.2.3 Leak Rate for the Large Pre-Existing Containment Leak Rate Case The third condition stipulates that in order to make the methodology in EPRI Technical Report No. 1009325, Revision 2, acceptable, the average leak rate for the pre-existing containment large leak rate accident case (i.e., accident case 3b) used by the licensees shall be 100 La instead of 35 La. As noted by the licensee in the table in Section 4.6.1 of Enclosure 2 of the LAR, the methodology in EPRI Technical Report No. 1009325, Revision 2-A, incorporated the use of 100 La as the average leak rate for the pre-existing containment large leakage rate accident case (accident case 3b), and this value has been used in the CNP plant specific risk assessment. Accordingly, the third condition is met.

3.3.2.4 Applicability if Containment Overpressure is Credited for ECCS Performance The fourth condition stipulates that in instances where containment over-pressure is relied upon for ECCS performance, a LAR is required to be submitted. In Section 4.6.1 of Enclosure 2 of the LAR, the licensee stated that CNP Units 1 and 2 do not rely on containment overpressure to assure adequate ECCS pump net positive suction head following design basis accidents.

Accordingly, the fourth condition is met.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Michigan State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or change the surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding as published in the Federal Register on May 27, 2014 (79 FR 30188).

Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

Based on the NRC staff's review of the licensee's submittal of March 7, 2014 (Reference 1), and supplemental letters dated September 30, 2014, December 16, 2014, January 15, 2015, and February 20, 2015 (References 2, 3, 4, and 5, respectively), and the regulatory and technical evaluation above, the NRC staff concludes that the licensee has effectively implemented an adequate containment leakage rate testing (ILRT and LLRT) program, CISI program, and supplemental inspections to periodically examine, monitor, and manage age-related degradation of the CNP Units 1 and 2 primary containment. The results of the past ILRTs, LLRTs, and the CISI program demonstrate acceptable performance of the CNP primary containment and demonstrate that the structural and leak-tight integrity of the primary containment structure is adequately managed. In addition, the staff finds that the licensee has addressed the NRC conditions to demonstrate acceptability of adopting NEI 94-01, Revision 3-A, without undue risk to public health and safety.

Therefore, the NRC staff concludes that it is acceptable to approve the proposed license amendment to revise TS 5.5.14, "Containment Leakage Rate Testing Program," to adopt NEI 94-01, Revision 3-A, as the implementation document, extend the performance-based Type A leakage test interval to 15 years, and extend the Type C leakage test interval to 75 months.

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1 Letter dated March 7, 2014, from Joel P. Gebbie, Indiana Michigan Power Company, to USNRC regarding License Amendment Request to Revise Donald C. Cook Nuclear Plant Units 1 and 2 Technical Specification Section 5.5.14, "Containment Leakage Rate Testing Program," Docket Nos. 50-315 and 50-316 (ADAMS Accession No. ML14071A435) 2 Letter dated September 30, 2014, from Joel P. Gebbie, Indiana Michigan Power Company, to USNRC regarding Response to Request for Additional Information Associated with License Amendment Request to Revise Donald C. Cook Nuclear Plant Units 1 and 2 Technical Specification Section 5.5.14, "Containment Leakage Rate Testing Program,"

Docket Nos. 50-315 and 50-316 (ADAMS Accession No. ML14275A454) 3 Letter dated December 16, 2014, from Joel P. Gebbie, Indiana Michigan Power Company, to USNRC regarding Supplemental Response to Request for Additional Information Associated with License Amendment Request to Revise Donald C. Cook Nuclear Plant Units 1 and 2 Technical Specification Section 5.5.14, "Containment Leakage Rate Testing Program," Docket Nos. 50-315 and 50-316 (ADAMS Accession No. ML14352A232) 4 Letter dated January 15, 2015, from Joel P. Gebbie, Indiana Michigan Power Company, to USNRC regarding Response to Request for Additional Information Associated with License Amendment Request to Revise Donald C. Cook Nuclear Plant Units 1 and 2 Technical Specification Section 5.5.14, "Containment Leakage Rate Testing Program," Docket Nos.

50-315 and 50-316 (ADAMS Accession No. ML15020A662) 5 Letter dated February 20, 2015, from Joel P. Gebbie, Indiana Michigan Power Company, to USNRC regarding Response to Request for Additional Information Associated with License Amendment Request to Revise Donald C. Cook Nuclear Plant Units 1 and 2 Technical Specification Section 5.5.14, "Containment Leakage Rate Testing Program," Docket Nos.

50-315 and 50-316 (ADAMS Accession No. ML15055A048) 6 Nuclear Energy Institute NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," July 2012 (ADAMS Accession No. ML12221A202) 7 Nuclear Energy Institute NEI 94-01, Revision 2-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," October 2008 (ADAMS Accession No. ML100620847) 8 NRC Final Safety Evaluation Report, "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 Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2, August 2007, 'Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," US Nuclear Regulatory Commission, Washington, DC, June 25, 2008 (ADAMS Accession No. ML081140105) 9 Electric Power Research Institute (EPRI) Technical Report No. 1018243, also identified as EPRI Technical Report 1009325, Revision 2-A, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals." Publicly available at www.epri.com by searching for "1018243."

Principal Contributors: F. Farzam, NRR M. Biro, NRR J. Bettle, NRR Date: March 30, 2015

ML15072A264

• via memo OFFICE LPL3-1/PM LPL3-1/PM LPL3-1/LA ORA/AP LA/BC DE/EMCB/BC NAME ADietrich MChawla MHenderson HHamzehee*

Tlupold*

DATE 03/13/2015 03/19/2015 03/18/2015 03/09/2015 01/20/2015 OFFICE DSS/SCVB/BC DSS/STSB/BC OGC NRR/LPL3-1 /BC NRR/LPL3-1 /PM NAME RDennig*

RElliott JWachutka DPelton ADietrich DATE 03/18/2015 03/24/2015 03/23/2015 03/30/2015 03/30/2015