ML120740081

Issuance of Amendment to Extend the Containment Type a Leak Rate Test Frequency to 15 Years

ML120740081
Person / Time
Site:	Palisades
Issue date:	04/23/2012
From:	Mahesh Chawla Plant Licensing Branch III
To:	Entergy Nuclear Operations
chawla M
References
TAC ME5997
Download: ML120740081 (32)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 23, 2012 Vice President, Operations Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043-9530 SUB~IECT: PALISADES NUCLEAR PLANT - ISSUANCE OF AMENDMENT TO EXTEND THE CONTAINMENT TYPE A LEAK RATE TEST FREQUENCY TO 15 YEARS (TAC NO. ME5997)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 247 to Renewed Facility Operating License No. DPR-20 for the Palisades Nuclear Plant (PNP). The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated April 6, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML110970616), supplemented by letter dated October 28,2011 (ADAMS Accession No. ML113010400).

The amendment revises TS 5.5.14, "Containment Leak Rate Testing Program," by replacing the reference to (RG) 1.163, "Performance-Based Containment Leak-Test Program," with a reference to Nuclear Energy Institute (NEI) topical report (TR) NEI 94-01, Revision 2-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR [Title 10 of the Code of Federal Regulations] Part 50, Appendix J," as the implementation document for the 10 CFR 50 Appendix J, Option B, performance-based containment leak rate testing program at the PNP. This amendment would allow PNP to extend its performance-based containment integrated leakage rate test (ILRT, or Type A test) interval up to 15 years. Accordingly, the licensee has also requested to extend its current Type A test interval from the current one-time approved 11.25 years to 15 years so that the next Type A test can be conducted by May 3, 2016, instead of the current due date of August 3, 2012.

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

Sincerely,

()~I~

Mahesh L. Chawla, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-255

Enclosures:

1. Amendment No. 247 to DPR-20

2. Safety Evaluation cc w/encls: Distribution via ListServ

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-255 PALISADES NUCLEAR PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 247 License No. DPR-20

1.

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

A.

The application for amendment by Entergy Nuclear Operations, Inc. (the licensee),

dated April 6, 2011, as supplemented by letter dated October 28, 2011, 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 10 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; 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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to the license amendment and Paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-20 is hereby amended to read as follows:

The Technical Specifications contained in Appendix A, as revised through Amendment No. 247, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION I)J~~I Istvan Frankl, Acting Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and Technical Specifications Date of Issuance: April 23, 2012

ATTACHMENT TO LICENSE AMENDMENT NO. 247 RENEWED FACILITY OPERATING LICENSE NO. DPR-20 DOCKET NO. 50-255 Replace the following page of the Renewed Facility Operating License No. DPR-20 with the attached revised page. The changed area is identified by a marginal line.

REMOVE INSERT Page 3 Page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT Page 5.0-18 Page 5.0-18

- 3 (1)

Pursuant to Section 104b of the Act. as amended. and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," (a) ENP to possess and use, and (b) ENO to possess, use and operate, the facility as a utilization facility at the designated location in Van Buren County, Michigan, in accordance with the procedures and limitation set forth in this license; (2)

ENO, pursuant to the Act and 10 CFR Parts 40 and 70, to receive, possess, and use source and special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3)

ENO, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use byproduct, source, and special nuclear material as sealed sources for reactor startup. reactor instrumentation, radiation monitoring equipment calibration, and fission detectors in amounts as required; (4)

ENO, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material for sample analysis or instrument calibration, or associated with radioactive apparatus or components; and (5)

ENO, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operations of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations in 10 CFR Chapter I and is subject to all applicable provisions of the Act; 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)

ENO is authorized to operate the facility at steady-state reactor core power levels not in excess of 2565.4 Megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2)

The Technical Specifications contained in Appendix A, as revised through Amendment No. 247, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

ENO shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the SERs dated 09/01/78, 03/19/80, 02/10/81, 05/26/83, 07/12/85, 01/29/86, 12/03/87, and 05/19/89 and subject to the following provisions:

Renewed License No. DPR-20 Amendment No. 247

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Safety Functions Determination Program (SFDP) (continued)

c.

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

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

5.5.14 Containment Leak 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 NEI 94-01, Revision 2-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated October 2008, with the following exceptions:

1.

Leakage rate testing is not necessary after opening the Emergency Escape Air Lock doors for post-test restoration or post-test adjustment of the air lock door seals. However, a seal contact check shall be performed instead.

Emergency Escape Airlock door opening, solely for the purpose of strong back removal and performance of the seal contact check, does not necessitate additional pressure testing.

2.

Leakage rate testing at Pa is not necessary after adjustment of the Personnel Air Lock door seals. However, a between-the-seals test shall be performed at 0 psig instead.

3.

Leakage rate testing frequency for the Containment 4 inch purge exhaust valves, the 8 inch purge exhaust valves, and the 12 inch air room supply valves may be extended up to 60 months based on com ponent performance.

b.

The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 54.2 psig. The containment design pressure is 55 psig.

c.

The maximum allowable containment leakage rate. La. at Pa, shall be 0.1 % of containment air weight per day.

Palisades Nuclear Plant 5.0-18 Amendment No. -'1-89. +94, ~,

~,~,244.

247

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 247 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-20 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT DOCKET NO. 50-255

1.0 INTRODUCTION

By letter dated April 6, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML110970616), supplemented by letter dated Ocober 28, 2011 (ADAMS Accession No. ML113010400), Entergy Nuclear Operations, Inc. (the licensee),

requested changes to the Technical Specifications (TSs) for the Palisades Nuclear Plant (Palisades or PNP). The proposed changes would revise TS 5.5.14, "Containment Leak Rate Testing Program," by replacing the reference to Regulatory Guide (RG) 1.163 (September 1995), "Performance-Based Containment Leak-Test Program," with a reference to Nuclear Energy Institute (NEI) topical report (TR) NEI 94-01, Revision 2-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR [Title 10 of the Code of Federal Regulations] Part 50, Appendix J," Revision 2-A, dated October 2008 (Reference 5.4), as the implementation document used by the licensee to develop the Palisades performance-based leakage testing program in accordance with Option B of 10 CFR Part 50, Appendix J.

In accordance with the guidance in NEI 94-01, Revision 2-A, the proposed change would permit the performance-based primary containment integrated leak rate testing (lLRT) interval to be extended from no longer than 10 years to no longer than 15 years provided acceptable performance is maintained. The license amendment request (LAR) was supplemented by information provided by the licensee by letter dated October 28, 2011 (Reference 5.2), in response to the staffs request for additional information (RAI), submitted via emails sent on August 29,2011 (ADAMS Accession No. ML112420181), and on September 2,2011 (ADAMS Accession No. ML112510336.)

The supplemental letter dated October 28, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on June 14, 2011 (76 FR 34766).

Enclosure

- 2

2.0 REGULATORY EVALUATION

Section 50.54(0) of 10 CFR requires 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 includes two options, Option A - Prescriptive Requirements, and Option B - Performance Based Requirements, either of which can be chosen for meeting the requirements of the Appendix.

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 at a periodic interval 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 integrity of the overall containment system as a barrier to fission product release.

Type A test is an overall, integrated leakage rate test (ILRT) of the containment structure.

NEI 94-01, Revision 0 specifies an initial test interval of 48 months, but allows an extended interval of 10 years, based upon two consecutive successful tests. There is also a provision for extending the test interval an additional 15 months in certain circumstances.

The testing requirements in Appendix J ensure that (a) leakage through these containments or systems and components penetrating these containments does not exceed allowable leakage rates specified in the TSs; and (b) integrity of the containment structure is maintained during its service life. PNP has voluntarily adopted and has been implementing Option B for meeting the requirements of Appendix J.

The leakage rate test results must not exceed the allowable leakage rate (La) with margin, as specified in the TSs. Option B also requires that a general visual inspection of the accessible interior and exterior surfaces of the containment system for structural deterioration, which may affect the containment leak-tight integrity, must 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 50 Appendix J, Option B, requires that the RG or other implementation document used by a licensee to develop a performance-based containment leak-test program must be included, by general reference, in the plant TSs. Further, 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 RG.

The implementation document that is currently referenced in the PNP TS 5.5.14, "Containment Leak Rate Testing Program," is RG 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995 with three exceptions. RG 1.163 (September 1995) endorsed NEI topical report NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance

- 3 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 to Appendix J to 10 CFR Part 50, 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 up to 10 years, based upon two consecutive successful tests.

By letter dated June 25, 200B (ADAMS Accession No. MLOB1140105), the NRC published a safety evaluation, with limitations and conditions, for NEI 94-01, Revision 2. In the Safety Evaluation (SE) the NRC concluded that NEI 94-01, Revision 2, describes an acceptable approach for implementing the optional performance-based requirements 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 limitations and conditions, noted in Section 4.0 of the SE.

The most recent two Type A tests at PNP were successfully completed in February 1991 and May 2001. Based on RG 1.163 (September 1995) as the implementing document, PNP is currently on a 10-year interval for Type A tests, and based on an approved one-time additional extension of 15 months (Reference 5.5) to the 1 O-year interval the next Type A test is due by August 3,2012, approximately 11.25 years since the last ILRT. The proposed amendment would allow the next ILRT for Palisades to be performed within 15 years from the last ILRT (i.e., May 3,2016). Therefore the next Type A test would be performed 45 months after the extension that was allowed on the 10-year interval. The successive Type A tests would have to be performed at 15-year intervals provided acceptable performance history is maintained.

The proposed TS change involves one new regulatory commitment which is included in of the LAR submittal (Reference 5.1). The commitment states that PNP will use the definition in Section 5.0 of NEI 94-01, Revision 2-A, for calculating the future Type A leakage rate on a continuing compliance basis following NRC approval of the LAR. The PNP Containment Leak Rate Testing Program will continue to comply with the requirements of 10 CFR Part 50, Appendix J.

3.0 TECHNICAL EVALUATION

3.1 Licensee's Proposed Changes The current TS 5.5.14 "Containment Leak Rate Testing Program," subsection a, currently states:

"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 of Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, except that the next Type A test performed after the May 3, 2001, Type A test shall be performed no later than August 3,2012, as modified by the following exceptions:"

-4 The licensees proposed request would modify TS 5.5.14 as follows:

"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 NEI 94-01, Revision 2-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated October 2008, with the following exceptions:"

3.2 Deterministic Considerations: Structural and Leak-Tight Integrity of the Containment As required by 10 CFR Part 50.54(0), the Palisades 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.

Currently, the Palisades Containment Leak Rate Testing Program is based on RG 1.163, which endorses NEI 94-01, Rev. O. This LAR proposes to revise the Palisades Containment Leak Rate Testing Program by implementing the guidance in NEI 94-01, Revision 2-A.

The proposed TS change does not involve any other changes to licensing commitments or acceptance criteria. Changing the reference from RG 1.163 to NEI 94-01, Revision 2-A is consistent with the NRC SE of the NEI guidance in NEI 94-01, Revision 2-A (Reference 5.3).

The Containment Leak Rate Testing Program requires the licensee to perform ILRT, also termed as a Type A test, and local leakage rate tests (LLRTs) termed as Type B and Type C tests. The Type A test measures the overall leakage rate of the primary reactor containment.

Type B tests are primarily intended to detect leakage paths and measure leakage rates for primary reactor containment penetrations. Type C tests are intended to measure containment isolation valve leakage.

The leak-tight integrity of the penetrations and isolation valves are verified through Type Band Type C LLRTs and the overall leak-tight integrity and structural integrity of the primary containment is verified through a Type A test (ILRT) as required by 10 CFR Part 50, Appendix J.

These tests are performed at the design-basis accident pressure. The testing frequency for Type B and Type C tests is not affected by the proposed amendment.

Type Band 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 Band C testing.

Palisades routinely performs various inspections and tests to assure primary containment integrity in addition to periodic Type A testing. These include Type Band C testing performed in accordance with 10 CFR Part 50, Appendix J, Option B; inspection activities performed as part of the American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME Code),

Section XI (Subsections IWE and IWL) inspection program; inspection of accessible interior and exterior surfaces of the containment system. The aggregate results of these tests and inspections provide a high degree of assurance of continued primary containment integrity.

- 5 The Palisades 10 CFR Part 50, Appendix J, Type B and Type C testing program evaluates electrical penetrations, airlocks, hatches, flanges, and valves within the scope of the program as required by 10 CFR Part 50 Appendix J, Option B, and TS 5.5.14. The Type Band C test program consists of local leak rate testing of penetrations with a resilient seal, double-gasketed manways, hatches and flanges, and containment isolation valves that serve as a barrier to the release of the post-accident primary containment atmosphere. The results of the test program are used to ensure that proper maintenance and repairs are made on the primary containment components over their service life. Type B and Type C testing provide a high degree of assurance that primary containment integrity is maintained.

The piping and ventilation penetrations are of the rigid welded type and are solidly anchored to the containment wall, thus precluding any requirement for expansion bellows.

The 10 CFR Part 50, Appendix J, Option B,Section III.A, states: "A general visual inspection of the accessible interior and exterior surfaces of the containment system for structural deterioration which may affect the containment leak-tight integrity must be conducted prior to each test, and at a periodic interval between tests based on the performance of the containment system." This inspection is also conducted during two other refueling outages before the next Type A test if the interval for the Type A test has been previously extended to 10 years, in order to allow for early discovery of structural deterioration. Effective September 1996, the NRC amended 10 CFR 50.55a to endorse Subsections IWE and IWL of the ASME Code,Section XI, 1992 Edition including 1992 Addenda. These subsections contain in-service inspection (lSI) and repair/replacement rules for Class MC (metal containment) and Class CC (concrete containment) components.

In addition to the IWElIWL lSI program inspections, the accessible interior and exterior surfaces of the containment system are visually inspected during each refueling outage. The visual inspection identifies evidence of structural problems that may affect either the containment structure leakage integrity or the performance of the Type A test.

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

3.2.1 Topical Report NEI 94-01! Revision 2-A NEI 94-01, Revision 2-A (Reference 5.4), is the NRC-accepted version of Revision 2 of the TR.

It incorporates the regulatory positions stated in RG 1.163 (September 1995), and includes provisions for extending Type A test (ILRT) intervals to up to 15 years. NEI 94-01, Revision 2-A, delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. This method uses industry performance data, plant-specific performance data, and risk inSights in determining the appropriate testing frequency. The guideline discusses the performance factors that licensees must consider in determining test intervals. While it does not provide the details on how to perform the tests, it references the national standard ANSIIANS 56.8-2002 (Reference 5.6) for detailed guidance for performing the tests.

-6 In the safety evaluation report (SER) dated June 25, 2008 (Reference 5.3), the NRC staff concluded that NEI 94-01, Revision 2-A, 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 reference 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 safety evaluation. Section 3.1 of the NRC SER provides the staff position on the adequacy of NEI 94-01, Revision 2, in addressing the performance-based Type A, Type B and Type C test frequencies. It also addresses the adequacy of pre-test inspections, procedures to be used after major modifications to the containment structure, deferral of tests beyond the 15 year interval, and the relation of containment lSI requirements mandated by 10 CFR 50.55a to the containment leak rate testing requirement.

NEI 94-01, Revision 2-A, also requires that a plant-specific risk impact assessment be performed using the approach and methodology described in TR-1009325, Revision 2-A, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," subject to conditions in Section 4.2 of the NRC SER (Reference 5.3), for a proposed extension of the ILRT interval to 15 years. The supporting plant-specific confirmatory risk assessment included in Attachment 4 of the LAR submittal (Reference 5.1) is discussed in Section 3.3 of this safety evaluation.

The licensee submitted the proposed TS change in accordance with 10 CFR Part 50 Appendix J, Option B,Section V.B.3, in order to change the implementation document referenced in TS 5.5.14, "Containment Leak Rate Testing Program."

3.2.2 Adoption of NEI 94-01, Revision 2-A, in TS 5.5.14 In the NRC SER dated June 25, 2008 (Reference 5.3), the staff concluded that the guidance in TR NEI 94-01, Revision 2-A, is acceptable for reference by licensees proposing to amend their TS in regards to containment leakage rate testing, subject to the six limitations and conditions noted as in Section 4.1 of the NRC SE for NEI 94-01, Revision 2-A. The NRC staff evaluated whether the licensee adequately addressed and satisfied these conditions in the LAR and supplemental information submittals (References 5.1 and 5.2), as discussed below.

a.

NRC Condition 1 NRC Condition1 states: "For calculating the Type A leakage rate, the licensee should use the definition in the NEI TR 94-01, Revision 2-A, in lieu of that in ANSIIANS-56.8 2002. (Refer to SE Section 3.1.1.1)."

In Table 4.0 of the LAR (Reference 5.1), the licensee stated that following NRC approval of this LAR, it will use the definition in Section 5.0 of NEI 94-01, Revision 2-A, for calculating the Type A leakage rate when future PNP Type A tests are performed.

Further, the licensee included this as a formal commitment in Attachment 5 "List of Regulatory Commitments" of the LAR applicable on a continuing compliance basis following NRC approval of the LAR. On the basis that the licensee has committed to comply with the definition in Sections 5.0, 9.1.1, and 9.2.3 of NEI 94-01, Revision 2-A, for calculating the Type A test performance leakage rate to demonstrate leakage integrity and to determine extended Type A test intervals, the NRC staff finds that the licensee has adequately addressed Condition 1 in its LAR.

- 7

b.

NRC Condition 2 NRC Condition 2 states: liThe licensee submits a schedule of containment inspections to be performed prior to and between Type A tests. (Refer to SE Section 3.1.1.3)."

NEI 94-01, Revision 2-A, Section 9.2.3.2 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 of the Type A test is extended to 15 years. NEI 94-01, Revision 2-A, Section 9.2.3.2 recommends performing the inspections in conjunction, or coordinated, with the ASME Code,Section XI, Subsection IWE/IWL required examinations.

Section 3.1.1.3 of the NRC SER states in part that, to avoid duplication or deletion of examinations, licensees using TR NEI 94-01, Revision 2-A, have to develop a schedule of containment inspections that satisfy both Section 9.2.3.2 of NEI 94-01, Revision 2-A, and ASME Code Section XI, Subsection IWE and IWL requirements.

The licensee discussed its schedule of containment visual inspections, with regard to the containment leak rate testing program, for a typical 15-year period between Type A tests in Section 4.3 of the LAR (Reference 5.1) and Attachment 2 of supplemental information submittal (Reference 5.2). The licensee stated that it typically conducts the general visual inspections in accordance with the PNP Containment lSI Plan, which implements the requirements of the ASME Code,Section XI, Subsections IWE (for metallic surfaces) and IWL (for concrete surfaces), as mandated by 10 CFR 50.55a(g)(4). The last Type A test was completed at PNP on May 3, 2001, during refueling outage 1 R15. The licensee stated that if the test interval was extended to 15 years, the next Type A test would be scheduled for refueling outage 1 R24 in 2015, approximately one year prior to the end of the 15-year interval.

With regard to the IWE inspections of containment steel liner surfaces, each ten-year lSI interval is divided into three inspection periods of approximately equal duration.

Subsection IWE requires 100 percent of the accessible containment liner surfaces to be subjected to general visual examination during each inspection period. The licensee stated that since a 15-year ILRT interval spans at least four lSI periods, the frequency of examinations per Subsection IWE (3 examinations over a 10-year interval) assures that at least three (may be four) general visual examinations of metallic components will be conducted between Type A tests and one scheduled immediately before the next Type A test, if the Type A test interval is extended to 15 years. The licensee concluded that, for the containment metallic liner, this satisfies the general visual requirements of Section 9.2.3.2 of NEI 94-01, Revision 2-A, and Condition 2 in the NRC SER.

Further, the licensee stated that visual examinations of accessible concrete containment surfaces are performed in accordance with Subsection IWL at a frequency of five years

(+/- 1 year). This results in at least three IWL examinations being performed during a 15-year Type A test interval, two of which would be between Type A tests. Since the last Type A test in 2001, two IWL examinations have been completed (2005 and 2010) and the next one is scheduled for 2015. The licensee stated that, in addition to the IWL

examinations, supplementary visual inspections of the accessible interior and exterior concrete surfaces of the containment structure are performed. These examinations are performed in sufficient detail to identify any evidence of deterioration which may affect the structural integrity or leak tightness of the containment building. The licensee stated that one such inspection was performed in 2009 and another inspection is scheduled to be performed prior to the Type A test in 2015. The licensee concluded that, together, these examinations assure that four general visual examinations of concrete containment surfaces (three between Type A tests and one prior to the next Type A test) will be conducted before the next Type A test, if the Type A test interval is extended to 15 years. Therefore, the requirements of Section 9.2.3.2 of NEI 94-01, Revision 2-A and Condition 2 in Section 4.1 of the NRC safety evaluation for NEI 94-01, Revision 2 would be met.

The licensee provided an illustration of an approximate inspection schedule, as shown in Table 1 (metallic surfaces) and Table 2 (concrete surfaces) below, for general visual examinations of accessible interior and exterior containment surfaces, representative of a typical 15-year period between Type A tests.

Table 1: Typical IWE 151 Schedule for Containment Metallic Liner Surfaces Inspection Period Start Period End Refuel Interval Date Date Outage 1

Oct 15, 1999 Feb 15, 2003 1R14 1R15 1

2 Feb 16, 2003 Jun 16, 2006 1R16 1R17 1R18 1

3 Jun 17,2006 Oct 15,2009 1

1 2

1 Oct 16, 2009 Feb 16. 2013 1R21 1R22 2

2 Feb17,2013 Jun 17,2016 1R23 1R24 2

3 Jun 18, 2016 Oct 16, 2019 1R25 1R26 1R27

-9 Table 2: PNP Typical Schedule of Concrete General Visual Examinations Calendar Yearl Outage Type A Test (ILRT)

Visual Examination of Accessible Exterior Concrete SurfacelType I

2001/1R1S X

X I Supplementary visual inspection 200S X/IWL Exam 2009 X I Supplementary visual inspection 2010 X/IWL Exam 201S/1R24 X

X I Supplementary visual inspection 201S X/lWLExam On the basis that the licensee's schedule of general visual examinations described above 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 staff finds that the licensee's inspection schedule plan, as detailed in the LAR and the supplemental information meets the general visual examination requirements in NEI 94-01, Revision 2-A, and 10 CFR Part SO Appendix J, Option B, and therefore, satisfies Condition 2 in the NRC SER for NEI 94-01, Revision 2-A.

c.

NRC Condition 3 NRC Condition 3 states: "The licensee addresses the areas of the containment structure potentially subjected to degradation. (Refer to SE Section 3.1.3)."

In Table 4.0 of the LAR (Reference S.1), the licensee stated that general visual examinations of accessible interior and exterior surfaces of the containment system for structural problems are typically conducted in accordance with the PNP Containment Inservice Inspection (CISI) Plan, which implements the requirements of the ASME Code,Section XI, Subsections IWE and IWL, as required by 10 CFR SO.SSa(g). The licensee stated that the applicable edition for its current second containment lSI interval is the 2004 Edition of ASME Code,Section XI. The PNP containment system does employ moisture barriers but does not employ expansion bellows on penetrations through the containment pressure retaining boundaries. The licensee stated that the PNP IWEIIWL program contains requirements to evaluate the acceptability of the inaccessible areas if such conditions were identified, in accordance with 10 CFR SO.SSa(b)(2)(ix){A) and 10 CFR SO.SSa(b)(2)(viii)(E). There are no primary containment surface areas that currently require an augmented examination in accordance with ASME Section XI, IWE-1240.

-10 The licensee stated that, consistent with the guidance in Section 9.2.3.3 of NEI 94-01, Revision 2-A, abnormal degradation of the containment structure, identified during the conduct of the IWE/IWL lSI program examinations or other inspections, are entered into the corrective action program for evaluation, to determine the cause of the degradation and to initiate appropriate corrective actions.

In the supplemental information provided in Attachment 2 of Reference 5.2 in response to the NRC staffs RAI, the licensee identified the specific areas of the PNP containment pressure boundary (both concrete and steel) that are inaccessible and susceptible to degradation. The licensee stated that in accordance with the PNP IWE program, the inaccessible portions of the containment liner plate are as follows:

(i) The portions of the liner plate below the concrete floor at the 590-foot elevation, (ii) Leak chase channels used to check the seams in the floor liner plate, which are embedded on the inaccessible side of the containment liner plate, (iii) Areas of the containment sump penetrations to containment liner connections which are embedded in concrete, and (iv) The portion of the fuel transfer tube which is encased in concrete or buried in the "gravel pit."

The licensee further stated that from the PNP IWL program, the inaccessible concrete areas are the portions of the concrete surface that are covered by the liner, foundation material, and backfill or are otherwise obstructed by adjacent structures, components, parts or appurtenances.

The licensee stated in the supplemental information in Reference 5.2 that there was one instance of an examination that identified a condition in an accessible area, above the moisture barrier of the containment, that suggested the potential for degradation in an inaccessible area. The licensee stated that prior to the implementation of the PNP IWEIIWL programs, an indication of general corrosion was discovered, during the 1998 refueling outage 1 R 13, above the moisture barrier around the containment circumference at the 590-foot elevation inside containment. This condition was documented and dispositioned. The corrective actions required investigation below the moisture barrier in the location containing indications as well as other locations.

Boroscope examinations determined that no unsatisfactory degradation existed, in the inaccessible area beneath the moisture barrier. The licensee further indicated that to date, there have been no additional instances in relation to the IWE or the IWL examinations where conditions were identified in accessible areas that would indicate the potential presence of degradation in the inaccessible areas, in accordance with 10 CFR 50.55a(b)(2)(ix)(A) and 10 CFR 50.55a(b)(2)(viii)(E). Thus, to date, PNP has not had a need to implement any new technologies to inspect the inaccessible areas.

The licensee added that it actively partiCipates in various nuclear utility owner groups, ASME Code committees, and with NEI, to maintain cognizance of ongoing developments within the nuclear industry. Industry operating experience is also continuously reviewed to determine its applicability to PNP. The licensee indicated that new commercially available technologies for examination of the inaccessible

- 11 degradation-susceptible areas of the containment in the future are explored and considered as part of these activities.

Section 3.1.3 of the NRC SER (Reference 5.3) for TR NEI 94-01, Revision 2-A, in part, states that licensees referencing NEI 94-01, Revision 2-A, in support of a request to amend their TS should also explore/consider such inaccessible degradation-susceptible areas in plant-specific inspections, using viable, commercially available nondestructive examination (NDE) methods (such as boroscopes, guided wave techniques, etc.- see Report ORNUNRC/L TR-02/02, "Inspection of Inaccessible Regions of Nuclear Power Plant Containment Metallic Pressure Boundaries," June 2002 (ADAMS Accession No. ML061230425), for recommendations to support plant-specific evaluations. The NRC staff's intent of this statement in the SER is that licensees should explore and consider NDE techniques such as those discussed in the reference or other state-of-the art methods for inspections of inaccessible degradation-susceptible areas of the containment pressure boundary, in a proactive manner, to support plant-specific evaluations of inaccessible areas, as these advanced technologies become commercially available and viable for implementation in practice in the future. While recognizing that these techniques may not be fully commercially viable at the present time, the NRC staff emphasizes that the issue related to inaccessible areas is especially important in light of several instances of significant through-wall containment liner corrosion degradations that have been identified in the last decade in US operating nuclear power plants, where the corrosion initiated at the inaccessible concrete-steel interface.

The information provided by the licensee identifies areas in the PNP containment that are inaccessible. The NRC staff observes that, in addition to the inaccessible areas identified by the licensee as discussed in previous paragraphs, the side of the containment liner plate backing to the concrete is one very important degradation susceptible inaccessible area where these NDE advanced technologies could be applied when they become viable, and should be included in the licensees inaccessible areas list in the IWE program. Except for the one dispositioned instance discussed previously, the licensee indicated that the PNP operating experience to date, has not identified any additional conditions that would indicate the presence or result in degradation of these inaccessible areas. Nevertheless, the licensee acknowledged that, as an active participant of tracking ongoing technology developments and industry operating experience, adjustments to inspection plans and the availability of new, commercially available technologies for examination of the inaccessible areas of the containment would be explored and considered as part of these activities. Therefore, the staff finds that the licensee has adequately addressed the intent of Condition 3 in its LAR.

d.

NRC Condition 4 NRC Condition 4 states: "The licensee addresses any tests and inspections performed following major modifications to the containment structure, as applicable. (Refer to SE Section 3.1.4)."

In Table 4.0 of the LAR, the licensee stated that the Entergy fleet design change process would address any testing and inspection requirements following future major structure,

- 12 as applicable, modifications to the containment structure. The license stated that its process provides a disciplined approach for determining the program and system interfaces associated with design changes. Specific questions are provided in this process pertaining to the ASME Containment In-Service Inspection Program, ASME Appendix J (Primary Containment Leak Rate Testing) Program, and ASME Section XI Repair/Replacement Program. These questions prompt the responsible engineer to consult with the applicable program owner for required actions including testing and inspections.

In the supplemental information provided in Reference 5.2 in response to the staffs RAI, the licensee stated that it had replaced the PNP steam generators during the 1990-91 refueling outage for which a construction opening was cut through the containment wall.

The construction opening was considered a major repair to the containment building.

The licensee stated that a Type A ILRT and a structural integrity test (SIT), were completed following containment restoration to show that the repairs to the containment adequately met the TSs leakage requirements. The Type A ILRT and SIT are documented in the PNP Final Safety Analysis Report, Sections 5.8.9.6.1 and 5.8.9.6.2.

The licensee added that the results of the SIT demonstrated that the containment was fully restored to the design condition existing prior to the steam generator replacement.

The licensee stated that, although currently there are no planned modifications, any unplanned modifications to the containment prior to the next scheduled Type A test (scheduled to be performed in the 2015 refueling outage 1R24 under this proposed change) would be subject to the special testing requirements of Section IV.A of 10 CFR Part 50, Appendix J. For minor modifications, leakage integrity of the affected pressure retaining areas should be verified by a local leak rate test.

The licensee clarified in its response that it understands the distinction between major and minor containment repairs and modifications, as described in Section 3.1.4 of the NRC SER (Reference 5.3) for TR NEI 94-01, Revision 2-A and in the statement of considerations for condition 10 CFR 50.55a(b)(2)(ix)(J) in the final rule published June 21, 2011 in the Federal Register 76 FR 36232-36270. The licensee indicated its understanding that, consistent with the discussion in the NRC's SER and in Section 9.2.4 of NEI 94-01, Revision 2-A, the minimum post-repair testing requirements following containment repairs or modifications as follows:

For major repairs or modifications (e.g., cutting of large openings for equipment removal/replacement, replacement of large penetrations, etc.): structural and leakage integrity of the restored containment should be verified by either a Type A test (ILRT) or a short duration structural test (as defined in Section 3.1.4 of NRC SER on NEI 94-01, Revision 2-A)

For minor repairs or modifications (e.g., items defined in IWE-5220): leakage integrity of the affected pressure retaining areas should be verified by a local leak rate test.

The NRC staffs intent of placing Condition 4, with reference to Section 9.2.4 of NEI 94-01, Revision 2-A, is to ensure that licensees clearly understand that following major containment modifications, such as those described in Section 3.1.4 of the NRC SER,

-13 the post-repair pressure testing performed must demonstrate both structural and leak-tight integrity of the repaired containment. Based on the information discussed above, and the post-repair pressure testing (Le., Type A test, structural integrity test) performed previously by the licensee in 1990-91 following the major containment modification for steam generator replacement, the staff finds that the licensee understands the NRC staffs 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 NRC SER for NEI 94-10, Revision 2-A. Therefore, the staff finds that the licensee has adequately addressed Condition 4 in its LAR.

e.

NRC Condition 5 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. (Refer to SE Section 3.1.1.2)."

The licensee stated in Table 4.0 of the LAR that it acknowledges and accepts the NRC staff position in Condition 5, as communicated to the nuclear industry in Regulatory Issue Summary (RIS) 2008-27, dated December 8, 2008.

The licensee has, thus, acknowledged and accepted the NRC staff position, with regard to extending the Type A test intervals beyond the approved upper bound limit of 15 years, in Condition 5 and clarified in 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." By this, the NRC staff finds that the licensee has confirmed its understanding that any extension of the Type A test interval beyond the upper-bound performance-based limit of 15 years should be infrequent and should be requested only for compelling reasons, and that the NRC staff will implement the position in RIS 2008-27 in reviewing such license amendment requests. Therefore, the staff finds that the licensee has adequately addressed Condition 5 in its LAR.

f.

NRC Condition 6 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 ILRT data."

The licensee stated in Table 4.0 of the LAR that this condition is not applicable since PNP is not licensed t01 0 CFR Part 52. The NRC staff finds that PNP is currently an operating reactor licensed to 10 CFR Part 50, and therefore, Condition 6 does not apply.

- 14 3.2.2.1 Conclusion for Licensee's Adoption of NEI 94-01, Revision 2-A Based on the above evaluation, the staff finds that the licensee has adequately addressed and satisfied the six conditions in Section 4.1 of the NRC SER for TR NEI 94-01, Revision 2-A, in its submittals (References 5.1 and 5.2). Therefore, the staff finds it acceptable for PNP to adopt TR NEI 94-01, Revision 2-A, as the implementation document in its TS 5.5.14, "Containment Leak Rate Testing Program."

3.2.3 Extension of Current Type A Test Interval from 11.25 Years to 15 Years 3.2.3.1 Description of the PNP Primary Containment System The Palisades containment structure consists of a post-tensioned, reinforced concrete cylinder and dome connected to and supported by a massive, reinforced concrete foundation slab. The entire interior surface of the containment structure is lined with 1/4-inch-thick welded ASTM A-442 steel plate to ensure a high degree of leak tightness. Numerous mechanical and electrical systems penetrate the containment wall through steel penetrations which are welded to the containment liner plate. The containment has an inside diameter of 116 ft, inside height of 189 ft, vertical wall thickness of 3.5 ft and dome thickness of 3 ft. Sufficient prestressing forces are applied to the cylinder and dome of the containment structure to more than balance the internal design pressure of 55 pounds per square inch gauge (psig). The post-tensioning system consists of: (a) three groups of 55 dome tendons oriented at 120 degrees to each other for a total of 165 tendons anchored at the vertical face of the dome ring girder; (b) 178 vertical tendons anchored at the top surface of the ring girder and at the bottom of the base slab; and (c) six groups of hoop tendons enclosing 120 degrees of arc for a total of 502 tendons anchored at the 6 vertical buttresses.

The containment structure completely encloses the primary coolant system and provides adequate biological shielding during both normal operation and accident situations. The containment structure is designed to ensure that leakage will not exceed 0.1 percent weight of containment air per day at a design pressure of 55 psig and a design temperature of 283°F.

The principal design basis for the structure is that it be capable of withstanding the internal pressure resulting from the design basis accident (DBA) with no loss of integrity.

The leak-tight integrity of the penetrations and isolation valves are verified through Type Band Type C LLRTs and the overall leak-tight integrity and structural integrity of the primary containment is verified through a Type A test (ILRT), as required by 10 CFR Part 50, Appendix J. These tests are performed at the peak calculated DBA pressure Pa, which is 53 psig for PNP. This peak calculated DBA pressure occurs during LOCA. The leakage rate testing requirements of 10 CFR Part 50 Appendix J, Option B (Type A, Type B and Type C Tests) and the Containment In-service Inspection (CISI) requirements mandated by 10 CFR Part 50.55a, together, ensure the continued leak-tight and structural integrity of the containment during its service life.

Under Option B, PNP is currently on an ILRT interval of 10 years, based on RG 1.163 (September 1995) as the implementation document, with a previously approved one-time extension to 11.25 years (Reference 5.5). By the current LAR (Reference 5.1), the licensee proposes to extend the current performance-based Type A test interval from 11.25 years to

-15 15 years by adopting TR NEI 94-01, Revision 2-A, as the implementation document in the TS.

This change would allow PNP to conduct the next Type A test by May 3,2016, in lieu of the current due date of August 3, 2012. The licensee justifies the proposed change by demonstrating adequate performance of the PNP containment based on plant-specific containment leakage testing program results and CISI program results and supported by a plant-specific risk assessment, consistent with the guidance in NEI 94-01, Revision 2-A. The LAR and supplemental information submittals (References 5.1 and 5.2) were reviewed and evaluated, as discussed in Sections 3.2.3.2 through 3.2.4, from the point of deterministic considerations with regard to containment structural and leak-tight integrity if the current ILRT interval is extended from 11.25 years to 15 years.

3.2.3.2 PNP Type A Test Performance History In LAR Section 4.1, the licensee discussed the results of all previous Type A tests conducted at PNP since May 1970 and the acceptance criteria specified in the TS at the time the tests were performed. Results of three recent tests and the leakage rate acceptance criteria are summarized in Table 3 below. The licensee stated that previous ILRT testing confirmed that the PNP containment structure leakage is acceptable with respect to the TS acceptance criterion for maximum allowable leakage, La, of 0.1 percent of containment air weight at the calculated peak internal pressure for the design-basis loss of coolant accident. The licensee stated that since the as-found results of the last two PNP Type A tests were less than 1.0 La, a test frequency of at least once per 15 years is justified in accordance with NEI 94-01, Revision 2-A.

TABLE 3: PNP Most Recent Type A Test Results

\\

Test Completion Date Test Pressure, psig Performance Leakage Rate

(% of containment air weight per day)

As-Found As-Left (S 0.1 % wtJday (S 0.075 % wtJday acceptable) (Note acceptable) (Note 1)

1)

May 3,2001 53.524 0.0140 0.0122 February 17, 1991 55.61 (Note 2) 0.070439 November 5, 1988 28.66 (reduced pressure

- Note 3) 0.0408 0.02617 Note 1: Per current TS 5-5.14.d.1, Type A test leakage rate acceptance criteria under Option B is s 1.0 La (for as-found) and S 0.75 La (for as-left), where La, the maximum allowable containment leakage rate at Pa, is 0.1% of containment air weight per day. Calculated peak containment internal pressure for LOCA, Pa = 53 psig, and containment design pressure, Pd =55 psig. Based on information in the LAR, for the tests performed until 1991 under Option A, the licensee previously used the acceptance criteria of S 0.75 La for both as-found and as-left conditions.

- 16 Note 2: Type A test was conducted at the end of the outage, prior to return to service, following repair (restoration) of the containment construction opening made for steam generator replacement, and was considered as an as-left post-modification preoperational test to demonstrate adequacy of repair.

Note 3: Reported results are corrected upward for calculated peak accident pressure, Pa.

The results in Table 3 indicates that previous three consecutive Type A tests at PNP. including the two most recent consecutive tests conducted at test pressure of Pa. were successful with containment performance leakage rate less than the maximum allowable containment leakage rate (La at Pa) of 0.1 percent containment air weight per day, at the calculated peak accident pressure of 53 psig. The staff finds. on the basis that the performance leakage rate for extending Type A test interval is determined by the licensee consistent with the definition in Sections 5.0, 9.1.1 and 9.2.3 of NEI 94-01. Revision 2-A (see NRC Condition 1 discussion in Section 3.3.1 of this SER input). the performance history of successful completion of two most recent consecutive periodic Type A tests supports extending the current ILRT interval to 15 years.

3.2.3.3 PNP Type B and Type C Tests The licensee described its Type B and Type C Testing Program in Section 4.2 of the LAR (Reference 5.1). The licensee stated that its Appendix J Type B and Type C testing program consists of local leak rate testing of penetrations (electrical and mechanical) with a resilient seal, double gasketed manways, hatches and flanges, and containment isolation valves that serve as a barrier to the release of the post-accident containment atmosphere. The licensee stated that PNP does not employ any expansion bellows.

Based on NE194-01, Revision 2-A, and TS 5.5.14, the combined Type B and Type C leakage rate acceptance criterion for Type B and Type C tests is less than 0.6 La for both as-found condition (on minimum pathway basis) as well as for as-left condition (on maximum pathway basis). The licensee stated that the values of La and 0.6La for PNP are 148.465 standard cubic centimeters per minute (sccm) and 89079 sccm. respectively. In Table 4.2-1 of the LAR, the licensee provided values of as-found minimum pathway combined Type B and Type C leakage test values since the last Type A test in 2001 through 2010. The Table indicates that there has been no outage since the last Type A test, in which the combined as-found minimum path leak rate from the Type B and Type C tests exceeded the acceptance criteria of 0.6La' During this period, the combined as-found leakage varied between 14 and 21 percent of 0.6La, except in two instances, where increased leakage of 37 percent (December 2001) and 54 percent (November 2004) of 0.6lawas reported. The largest contribution of the as-found leakage in these two instances was attributed to a specific but different penetration in each case. The excessive leakage at these penetrations were fixed, retested and placed on an increased test frequency (30-month) in accordance with the lLRT program. The most recent value of as-found combined Type B and Type C leakage rate for the PNP containment determined in October 2010 was 18,410 sccm, which is 20.7 percent of 0.6La. From Table 4.2-2 of the LAR, the latest as-left combined leakage rate is 18527.8 sccm, which is 21 percent of the acceptance limit of 0.6La. These results indicate that the combined leakage from the Type B and Type C tests has been consistently maintained well below the acceptance criteria.

- 17 In Table 4.2-2 of the LAR, the licensee identified the 42 PNP pressure boundary penetrations subject to Type B and Type C testing, their current test frequencies that were established under Option B based on performance, the last test date, the date for the next test, and the latest as left leakage rate. The Table also provided information on penetrations that have exceeded their administrative leakage limit in the past, indicated significantly increased leakage rates, or have been replaced, and how the performance-based test interval was adjusted. The Table indicates that of the 42 penetrations, 5 are subject to Type B testing and 37 to Type C testing. The Table indicates that currently, of the total of 42 penetrations, 1 (Type B penetration) is on the 120-month performance-based interval, 29 on a 60-month interval, 2 on a 48-month interval, and 10 (3 of which are personnel airlock, escape air lock and equipment hatch required to be tested every outage) are on a 30-month interval. It should be noted that of the 37 Type C penetrations, 29 are on the maximum performance-based interval of 60 months, which indicates generally good performance. During the most-recent testing in the 2010 refueling outage, 2 penetrations exceeded the administrative leakage rate and one penetration showed a significant increase in leakage but within the administrative limit. The licensee stated that the performance-based interval of these three penetrations was reduced to the base test interval of 30 months. Two penetrations continue to remain on a 48-month interval even though they meet the requirements for a 60-month interval. The data in the Table also indicate that Type Band Type C tests are scheduled in a staggered manner such that approximately equal numbers of components are tested during each refueling outage. Based on the information discussed above, the staff finds that the licensee has appropriately taken corrective actions and has adjusted test schedules consistent with its Appendix J, Option B program, for cases of as-found test failures.

The licensee stated that industry experience has shown that the Type Band C tests can identify the vast majority (over 95 percent) of all potential containment leakage paths. The licensee stated that this LAR would adopt the guidance in NEI 94-01, Revision 2-A, in place of NEI 9-01, Revision 0, but otherwise does not affect the scope, performance, or scheduling 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 containment integrity is maintained.

Based on the information discussed above, the staff finds that there is reasonable assurance that the licensee is appropriately implementing its Type B and Type C Testing program under Option B, in a rational and systematic manner that is consistent with the implementation document in the TS, and will continue to do so, in accordance with NEI 94-01, Revision 2-A, if the current ILRT interval is extended to 15 years. Thus, the staff finds that the integrity of the containment pressure boundary penetrations (including access hatches and airlocks) and isolation valves are effectively monitored through Type B and Type C testing, as required by 10 CFR Part 50, Appendix J and the implementation document referenced in the PNP TS.

3.2.4 Containment In-Service Inspection Program (ASME Section XI, Subsections IWE and IWL)

In Section 3.2.2 of the LAR submittal (Reference 5.1), the licensee stated that the PNP Containment Inservice Inspection (CISI) program (also referred to as the IWEIIWL program) which implements the requirements of ASME Code,Section XI, Subsections IWE and IWL, pursuant to 10 CFR 50.55a(g)(4), The applicable code edition/addenda for the second (current) 10-year interval of the program is the 2004 edition.

- 18 In Section 4.3.1 of the tAR, the licensee summarized highlights of the results of the most recent IWE examinations of the containment metallic liner performed during refueling outages 1 R18 (2006) and 1 R20 (2009). The licensee stated that the next IWE examination is scheduled for 1 R22 (2012). The licensee stated that IWE examinations performed per TS Surveillance Procedure RT-142, "Containment Inservice Inspection-Metal Liner," documented several indications in the PNP corrective action program (CAP) that were categorized as surface corrosion. The corrosion was clarified as not excessive and validated by performance of ultrasonic test examinations in several areas that were representative of all corroded areas in containment. The minimum liner thickness reported for these areas was 0.234 inches (in areas with nominal 0.250-inch thickness) and 0.485 inches (in areas with nominal 0.500-inch thickness). The examinations also identified a small area of missing moisture barrier, which was subsequently replaced and successfully re-inspected. The licensee concluded that the visual indications found were considered cosmetic with no areas of suspect damage or deterioration that would impact the structural integrity or leak tightness of the containment liner.

The RT-142 examinations of the containment liner plate were successfully completed and met the applicable code criteria for "Acceptance by Examination". The licensee stated that there are no primary containment surface areas that require augmented examination in accordance with ASME Section XI, IWE-1240.

In Section 4.3.2 of the tAR, the licensee provided the IWL examination results of the containment concrete visual inspections completed in 2000, 2005 and 2010; and tendon surveillance inspections completed in 2002 and 2008. The licensee stated that the next concrete inspection is scheduled for 2015 and the next tendon inspection is scheduled for 2012.

The licensee stated that IWL concrete visual examinations were performed during the summer of 2000 and June, 2005 under TS surveillance procedure FT-7 "Containment Visual Inspection,"

various minor recordable indications were observed. Examples of these indications included:

(a) tendon grease at various tendon buttresses, (b) actual grease leakage was observed at tendon caps in the tendon tunnel and on the containment dome, and (c) concrete pop-outs, spalls, cracks, and indications described in visual examination procedure. Some of the concrete pop-outs exposed rebar near four vertical tendons. These observations were entered into the PNP CAP. The licensee stated that evaluation of these identified observations of degradation found them to be minor affecting only the outer surfacial portions of the concrete structure and were considered to be cosmetic in nature. Further, historical information documented in the PNP CAP has indicated that grease leakage has not resulted in tendon wire corrosion.

The licensee stated that concrete visual examinations performed during summer 2010 under TS Surveillance Procedure FT -7, also recorded concrete pop-out. spalls, and cracks. All newly discovered indications were documented in accordance with inspection code requirements, reviewed by the Responsible Engineer and determined to support continued containment operability. No conditions in accessible areas indicated the presence of or could reasonably result in degradation of inaccessible areas. The licensee stated that it determined that the containment was fully capable of performing its protective and fission boundary functions.

The licensee stated that the 30th-year IWL tendon surveillance activities completed in January 2003 resulted in inspection observations. findings and corrective actions as follows.

Grease replacement in the sheathing of one dome tendon and four vertical tendons, which varied from 8.8 gallons (10.5 percent) to 18.3 gallons (32 percent), exceeded the acceptance criteria. However, each of these tendons met the criteria for lift-off force measurement,

- 19 anchorage hardware and surrounding concrete. Tendon wire surfaces were fully covered with a protective grease coating. All these tendon sheaths were refilled by injecting new grease and these tendons were determined to be fully operable. Also, grease leakage was discovered at the main gaskets at the top end (shop end) on the containment dome for nine vertical tendons.

The licensee determined that heating of the grease, following the filling of the grease cans during the steam generator replacement project, which occurred during a cold weather period in 1991, expanded the grease and pushed it by the main gasket. As part of the tendon surveillance project, the main gaskets were replaced and grease leakage from the subject cans stopped. The quantity of grease replacement was sufficient to cover tendon end anchorage hardware but an air pocket was left in the upper portion of the can to allow grease expansion and contraction. The protective grease layer on the tendons was not compromised by the small observed leakage.

The licensee also discovered three missing button heads on vertical tendon V-30, at the field end. The licensee noted that one of the missing button ends was previously recorded during plant construction. The cause of the two additional failed button heads appeared to be fabrication flaws inserted during the button heading process as evidenced by the lack of button heads in the removed grease. However, the licensee stated that there was no visible sign of deterioration at the end of the individual tendon wires, and vertical tendon V-30 met all the other applicable tests and inspection criteria. The licensee stated that it also discovered a single protruding wire at the shop end of dome tendon 03-22 and attributed it to a break somewhere along the length of the wire as evidenced by all button heads being in place at the field end.

Efforts to remove the wire were unsuccessful making it impossible to determine the cause of failure. There was no visible sign of deterioration at the ends of the wires. Tendon force measurement testing was not performed on dome tendon 03-22 due to obstructions. However, all other test and inspection acceptance criteria were met for grease coating, sampling and loss, inspection for water, anchorage corrosion, and concrete inspection.

The licensee stated that water infiltration has been documented during previous tendon surveillance at PNP. The cause of water infiltration has been traced to degraded grease can gaskets and migration through concrete cold joints and tendon sheathing. Surveillance activities discovered 20 ounces of free water at the shop end of dome tendon 01-38.

Additionally, the grease sample testing for this tendon indicated chemically combined water at 11 percent at the shop end, only. The licensee stated that. in order to fully determine tendon condition, tendon lift-off force measurement tests and visual exams were performed. Force measurements tests were satisfactory and inspections did not discover any degradation of anchorage components. On the basis of this information, the licensee concluded that the presence of free water or chemically combined water in the grease was insufficient to cause corrosion or cracking of the anchorage components. Oome tendon 01-38 was refilled by injecting new grease. In summary, the licensee concluded that following the 30th-year tendon surveillance, the containment structural integrity surveillance program had demonstrated continued containment operability and that the containment post-tensioning system had not experienced abnormal degradation.

The licensee stated that the 35th-year IWL tendon surveillance activities completed in September 2008 resulted in inspection observations. findings and corrective actions as follows.

Sheathing filler (grease) samples were tested and found to have acceptable levels of water soluble ions (chlorides. nitrates and sulfides) and neutralization numbers. The top end of one

- 20 vertical tendon and one end of a dome tendon were found respectively with water content of 14 percent and 18 percent by weight, which was above the acceptance limit of 10 percent. Both these tendon ends had acceptable grease coverage and corrosion inspection results. Two dome tendon (01-38 and 01-36) ends contained free water, of 23 ounces and 1 ounce, during removal of the grease cap. pH testing of a water sample from 01-38 returned an acceptable pH level of 12.80. The licensee stated that corrosion levels were found acceptable on all tendon ends and no cracks were found on any anchorage components. Cracks in the concrete surrounding the bearing plates were all within allowable width of less than 0.010 inch. The licensee stated the measured lift-off forces in all the sampled surveillance tendons were found to be greater than the acceptance limit for individual tendons of 95 percent of the corresponding predicted force. The tendons that were detensioned were retensioned with acceptable elongations and acceptable force levels. The licensee added that all test wires removed from de-tensioned tendons were found to have acceptable corrosion levels and all tendon test wire samples tested had acceptable diameter, yield stress, ultimate stress and elongation results.

Further, all tendons were resealed and re-greased to acceptable levels. The licensee stated that a comparison of as-found tendon force levels to the original force levels was made in an effort to detect any evidence of system degradation. The licensee stated that the amount of loss of prestress force since the original installation is comparable to the losses of other plants of this age and does not indicate any evidence of system degradation. Based on the data gathered during the 35th-year containment IWL tendon surveillance, the licensee concluded that no abnormal degradation of the post-tensioning system had occurred on the PNP containment structure.

Based on the results of the recent IWEIIWL inspections discussed in the previous paragraphs, the staff finds that there has not been evidence to date of significant degradation of the PNP containment structure, and the degradations noted have been entered into the PNP corrective action program, and appropriately managed and/or corrected. Based on the above evaluation, the staff finds that the licensee is adequately implementing its containment inservice inspection program to monitor and manage age-related degradation of the PNP containment structure.

The results of the inspections, to date, indicate that the structural and leak-tight integrity of the containment have been appropriately monitored and managed and will continue to be monitored and managed, if the current ILRT interval is extended from 11.25 years to 15 years, in accordance with NEI 94-01, Revision 2-A.

3.3 PROABILISTIC RISK ASSESSMENT 3.3.1

Background

Section 9.2.3.1 of NEI 94-01, Revision 2-A (October 2008), "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," states that plant-specific confirmatory analyses of the risk associated with ILRT interval extensions are required when extending the interval beyond ten years. Section 9.2.3.4 of NEI 94-01 states that the assessment should be performed using the approach and methodology described in Electric Power Research Institute (EPRI) Topical Report (TR) 1018243, Revision 2-A (October 2008),

"Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals." 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.

- 21 3.3.2 Plant-Specific Risk Evaluation The licensee performed a risk impact assessment for extending the Type A test interval from 10 to 15 years. The risk assessment was provided in the application for license amendment dated April 6, 2011. Additional information was provided by the licensee in its letter dated October 28, 2011, in response to the NRC staff's RAI. In performing the risk assessment, the licensee considered the guidelines of NEI 94-01, Revision 2-A, the methodology used in EPRI TR-1018243, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals,"

October 2008, and NRC RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," dated November 2002.

NUREG-1493, "Performance-Based Containment Leak-Test Program," dated September 1995, provided the technical basis 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 this basis, the industry undertook a similar study; the results of that study are documented in EPRI TR-104285.

The EPRI study used an analytical approach similar to that presented in NUREG-1493 for evaluating the incremental risk associated with increasing the interval for Type A tests. The Appendix J, Option A, requirements that were in effect for PNP early in the plant's life required a Type A test frequency of three tests in 10 years. The EPRI study estimated that relaxing the test frequency from three tests in 10 years to one test in 10 years would increase the average time that a leak, that was detectable only by a Type A test, goes undetected from 18 to 60 months. Since Type A tests only detect about 3 percent of leaks (the rest are identified during local leak rate tests, based on industry leakage rate data gathered from 1987 to 1993),

this results in a 10 percent increase in the overall probability of pre-existing containment leakage. The risk contribution of pre-existing leakage for the pressurized-water reactor and boiling-water reactor representative plants in the EPRI study confirmed the NUREG-1493 conclusion that a reduction in the frequency of Type A tests from three tests in 10 years to one test in 20 years leads to an "imperceptible" increase in risk that is on the order of 0.2 percent and a fraction of one person roentgen equivalent man (rem) per year in increased public dose.

The licensee quantified the risk from sequences that have the potential to result in large releases if a pre-existing leak was present. Since the Option B rulemaking was completed in 1995, the NRC staff has issued RG 1.174 on the use of probabilistic risk assessment (PRA) in evaluating risk-informed changes to a plant's licensing basis. The licensee has proposed using RG 1.174 guidance to assess the acceptability of extending the Type A test interval beyond that established during the Option B rulemaking.

RG 1.174 states that a PRA used in risk-informed regulation should be performed in a manner that is consistent with accepted practices. In NRC Regulatory Issue Summary (RIS) 2007-06, "Regulatory Guide 1.200 Implementation," dated March 22, 2007 (ADAMS Accession No. ML070650428), the NRC clarified that for all risk-informed applications received after December 2007, the NRC staff will use Revision 1 of RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

dated January 2007 (ADAMS Accession No. ML070240001), to determine whether the technical

- 22 adequacy of the PRA used to support a submittal is consistent with accepted practices.

Revision 2 of RG 1.200 will be used for all risk-informed applications received after March 2010.

In the Final Safety Evaluation for NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2 (ADAMS Accession No. ML081140105), the NRC staff states that Capability Category I of the ASME PRA Standard shall be applied as the standard for assessing PRA quality for ILRT extension applications, since approximate values of core damage frequency (CDF) and large early release frequency (LERF) and their contribution among release categories are sufficient to support the evaluation of changes to ILRT frequencies.

3.3.2.1 Technical Adequacy of the PRA The licensee's April 6, 2011, license amendment request addresses the technical adequacy of the PRA that forms the basis for the subject risk assessment. As provided in the response to the NRC staff's RAI (Reference 5.2), the PNP PRA internal events model is updated to meet ASME PRA Standard RA-S-2008a Capability Category II and RG 1.200, Revision 2 which is consistent with the requirements in the later standard RA-Sa-2009. An industry peer review team (Combustion Engineering Owners Group) reviewed the PNP PRA model in October 2009.

As part of the ILRT extension application and in response to NRC staff's RAI, the licensee confirmed that evaluations for all 38 findings that are not related to flooding and did not meet capability category I and II are provided in Appendix A, Table A.2.3-1, of the LAR The NRC staff reviewed this information and has no objection to the conclusions in the licensee's assessment. The impact of Revision 2 of RG 1.200 was considered for this assessment. The licensee performed a bounding analysis for the external events contributors and the results did not significantly increase the risk associated with this license amendment. Given that the licensee has evaluated its PRA against RG 1.200 and the ASME PRA Standard, evaluated all of the findings developed during the reviews of its PRA for applicability to the ILRT extension, and determined that any unresolved issues would not impact the conclusions of the ILRT risk assessment, the NRC staff concludes that the current PNP PRA model is of sufficient technical quality to support the evaluation of changes to ILRT frequencies.

3.3.2.2 Estimated Risk Increase RG 1.174 provides risk-acceptance guidelines for assessing the increases in CDF and LERF for risk-informed license amendment requests. Since the Type A test does not impact CDF, the relevant criterion is the change in LERF. The licensee has estimated the change in LERF for the proposed amendment based on the cumulative change from the original frequency of three tests in a 10-year interval, the current test interval of 10 years, and the proposed testing interval of 15 years. RG 1.174 also discusses defense-in-depth. The licensee estimated the change in the conditional containment failure probability for the proposed amendment and judged it to be insignificant and reflecting sufficient defense-in-depth.

The licensee comparisons of risk are based on a change in test frequency from three tests in 10 years (the test frequency under Appendix J, Option A) to one test in 15 years. This bounds the impact of extending the test frequency from one test in 10 years to one test in 15 years. The following conclusions can be drawn from the licensee's analysis associated with extending the Type A test frequency:

- 23

1. Given the change from the current 1 in 1 O-year test frequency to a one in 15-year test frequency, the increase in the total population dose is estimated to be 0.068 person-rem per year. This increase is consistent with the value estimated in NUREG-1493, where it was concluded that a reduction in the frequency of tests from three in 10 years to one in 20 years leads to an uimperceptible" increase in risk that is on the order of a fraction of one person-rem per year in increased public dose. This increase in the total integrated plant risk for the proposed change is considered small and supportive of the proposed change.

2. The increase in LERF resulting from a change in the Type A test frequency from the 3 in 10 years to one in 15 years is estimated to be about 2.44 x 10.7 per year, based on the plant-specific internal events PRA, and about 4.88 x 10.7 per year, when external events are included.

3. Guidance in RG 1.174 defines very small changes in LERF as below 10*7/yr and small changes in LERF below 10*s/yr. The NRC staff concludes that increasing the Type A interval to 15 years results in only a small change in LERF and is consistent with the acceptance guidelines of RG 1.174.

4. RG 1.174 also discusses the need to show that the proposed change is consistent with the defense-in-depth philosophy. Consistency with the defense-in-depth philosophy is maintained if a reasonable balance is preserved between prevention of core damage, prevention of containment failure, and consequence mitigation. The licensee estimates the change in the conditional containment failure probability to be an increase of less than 1 percentage point for the cumulative change of going from a test frequency of 3 in 10 years to one in 15 years. The NRC staff finds that the defense-in-depth philosophy is maintained based on the small magnitude of the change in the conditional containment failure probability for the proposed amendment.

Based on these conclusions, the NRC staff concludes that the increase in projected risk due to the proposed change is within the acceptance guidelines, while maintaining the defense-in-depth philosophy of RG 1.174, and is, therefore, acceptable.

4.0

SUMMARY

In summary, based on the regulatory and technical evaluations above, the staff finds that the licensee has adequately implemented its Containment Leakage Rate Testing consisting of ILRT and Local Leak Rate Testing (LLRT), containment inservice inspection (CISI) and supplementary inspections. The results of the recent ILRTs, LLRTs and the CISI programs demonstrate acceptable performance of the PNP containment and demonstrate that the structural and leak-tight integrity of the containment structure is adequately managed and will continue to be periodically monitored and managed by the ILRTs, LLRT and CISI programs.

The staff finds that the licensee has addressed the NRC conditions to demonstrate acceptability of adopting TR NEI 94-01, Revision 2-A, without undue risk to public health and safety.

Therefore, the staff concludes that it is acceptable to approve the proposed license amendment in Reference 5.1 for PNP to: (1) revise TS 5.5.14, "Containment Leak Rate Testing Program," to

- 24 adopt NEI 94-01, Revision 2-A, as the implementation document, and (ii) extend the current performance-based Type A test interval from 11.25 years to up to 15 years.

5.0 REFERENCES

5.1 Letter No. PNP 2011-018 dated April 6, 2011, from Thomas P. Kirwin, Entergy Nuclear Operations Inc - Palisades Nuclear Plant, to USNRC regarding License Amendment Request to Extend Containment Type A Leak Rate Test Frequency to 15 years (ADAMS Accession No. ML110970616) 5.2 Letter No. PNP 2011-067 dated October 28,2011, from Anthony J. Vitale, Entergy Nuclear Operations Inc - Palisades Nuclear Plant, to USNRC regarding Response to Request for Additional Information - Palisades LAR to Extend Containment Type A Leak Rate Test Frequency to 15 years (ADAMS Accession No. ML113010400) 5.3 NRC Final Safety Evaluation Report, "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report 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 Test Intervals," US Nuclear Regulatory Commission, Washington, DC, June 25, 2008 (ADAMS Accession No. ML081140105).

5.4 Nuclear Energy Institute Topical Report 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).

5.5 Letter dated August 23, 2010, from USNRC to Vice President, Operations, Entergy Nuclear Operations, Inc., regarding Palisades Nuclear Plant - Issuance of Amendment No. 240 RE: One-time extension of the Integrated Leak Rate Test Interval (TAC No.

ME2122) (ADAMS Accession No. ML102090137).

5.6 American Nuclear Society, "Containment System Leakage Testing Requirements,"

ANSI/ANS 56.8-2002, LaGrange Park, Illinois.

6.0 STATE CONSULTATION

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

7.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes the surveillance requirements. The staff has determined that the amendment involves 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 amendment involves no significant hazards consideration and there has been no public

- 25 comment on such finding (76 FR 34766). Accordingly, the amendment meets 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 amendment.

8.0 CONCLUSION

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) such activities will be conducted in compliance with the Commission'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.

Principal Contributors: Brian Lee George Thomas Jerry Dozier Date: April 23, 2012

ML120740081 OFFICE LPL3-1/PM LPL3-1/LA SCVB/BC EMCB/BC NRR/DRAlBC NAME MChawla BTuily RDennig MMurphy DHarrison DATE 03/22/12 03/22/12 03/26/12 03/27/12 03/27/12 OFFICE ITSB/BC OGC(NLO w/comments)

LPL3-1/BC(A)

LPL3-1/PM NAME RElliot LSubin IFrankl MChawla DATE 03/29/12 04/05/12 04123112 04/23/12