ML15078A058

Issuance of Amendment License Amendment Request to Extend Containment Leakage Rate Test Frequency (TACs MF3985 and MF3986)

ML15078A058
Person / Time
Site:	Beaver Valley
Issue date:	04/08/2015
From:	Taylor Lamb Plant Licensing Branch 1
To:	Emily Larson FirstEnergy Nuclear Operating Co
Lamb T, DORL/LPLI-2, 415-7128
References
TAC MF3985, TAC MF3986
Download: ML15078A058 (36)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 8, 2015 Mr. Eric A. Larson, Site Vice President FirstEnergy Nuclear Operating Company Beaver Valley Power Station Mail Stop A-BV-SEB1 P.O. Box 4, Route 168 Shippingport, PA 15077

SUBJECT:

BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENT RE: LICENSE AMENDMENT REQUEST TO EXTEND CONTAINMENT LEAKAGE RATE TEST FREQUENCY (TAC NOS. MF3985 AND MF3986)

Dear Mr. Larson:

The Commission has issued the enclosed Amendment No. 293 to Renewed Facility Operating License No. DPR-66 for the Beaver Valley Power Station, Unit No. 1 (BVPS-1 ), and Amendment No. 180 to Renewed Facility Operating License No. NPF-73 for the Beaver Valley Power Station, Unit No. 2 (BVPS-2). These amendments consist of changes to the technical specifications (TS) in response to your application dated April 16, 2014, as supplemented by letters dated November 4, 2014, and March 23, 2015.

The amendment modifies BVPS TS 5.5.12, "Containment Leakage Rate Testing Program," Item a, by deleting reference to the BVPS-1 exemption letter dated December 5, 1984, and requiring compliance with Nuclear Energy Institute (NEI) topical report NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR [Title 10 of the Code of Federal Regulations] Part 50 Appendix J," instead of Regulatory Guide 1.163, "Performance Based Containment Leak Test Program," including listed exceptions. This change will allow BVPS-1 and BVPS-2 to extend the Type A reactor containment test testing interval, required by 10 CFR Part 50 Appendix J, from one test in 10 years to one test in 15 years, and extension of the Type C test interval up to 75 months, with a permissible extension period of 9 months (total of 84 months) for non-routine emergent conditions.

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

Ov(j~

Taylor A. La b, Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-334 and 50-412

Enclosures:

1. Amendment No. 293 to DPR-66

2. Amendment No. 180 to NPF-73

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FIRSTENERGY NUCLEAR OPERATING COMPANY FIRSTENERGY NUCLEAR GENERATION, LLC DOCKET NO. 50-334 BEAVER VALLEY POWER STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 293 Renewed License No. DPR-66

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

A. The application for amendment by FirstEnergy Nuclear Operating Company, (FENOC)* acting on its own behalf and as agent for FirstEnergy Nuclear Generation, LLC (the licensees), dated April 16, 2014, as supplemented by letters dated November 4, 2014, and March 23, 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 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 regulations of the Commission; 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-66 is hereby amended to read as follows:

• FENOC is authorized to act as agent for FirstEnergy Nuclear Generation, LLC, and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 293, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION Douglas A. Broaddus, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: April 8, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 293 RENEWED FACILITY OPERATING LICENSE NO. DPR-66 DOCKET NO. 50-334 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page 3 Page 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 area of change.

Remove 5.5-19 5.5-19

(3) FE NOC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) FENOC, 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 without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) FE NOC, 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 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 1:

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 FENOC is authorized to operate the facility at a steady state reactor core power level of 2900 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 2 9 3, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Auxiliary River Water System (Deleted by Amendment No. 8)

Amendment No. 2 9 3 Beaver Valley Unit 1 Renewed Operating License DPR-66

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP) (continued)

c. 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.

5.5.12 Containment Leakage Rate Testing Program

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. For Unit 1, exemptions to Appendix J of 10 CFR 50 are dated November 19, 1984, and July 26, 1995. For Unit 2, exemptions to Appendix J of 10 CFR 50 are as stated in the Operating License. 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 ," and conditions and limitations specified in NEI 94-01, Revision 2-A.

b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 43.1 psig (for Unit 1) and 44.8 psig (for Unit 2).

c. The maximum allowable containment leakage rate, La, at Pa, shall be 0.10% of containment air weight per day.

d. Leakage rate acceptance criteria are:

1. Containment leakage rate acceptance criterion is ::; 1.0 La. However, during the first unit startup prior to MODE 4 entry following testing in accordance with this program, the leakage rate acceptance criteria are

< 0.60 La for the Type Band C tests and::; 0.75 La for Type A tests.

2. Air lock testing acceptance criteria are:

a) Overall air lock leakage rate is::; 0.05 La when tested at;::: Pa.

Beaver Valley Units 1 and 2 5.5 -19 Amendments 293 /180

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FIRSTENERGY NUCLEAR OPERATING COMPANY FIRSTENERGY NUCLEAR GENERATION LLC OHIO EDISON COMPANY THE TOLEDO EDISON COMPANY DOCKET NO. 50-412 BEAVER VALLEY POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 180 Renewed License No. NPF-73

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

A. The application for amendment by FirstEnergy Nuclear Operating Company (FENOC)* acting on its own behalf and as agent for FirstEnergy Nuclear Generation, LLC, Ohio Edison Company, and The Toledo Edison Company (the licensees), dated April 16, 2014, as supplemented by letters dated November 4, 2014, and March 23, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 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 rules and 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.

• FENOC is authorized to act as agent for FirstEnergy Nuclear Generation, LLC, Ohio Edison Company, and The Toledo Edison Company and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.

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. NPF-73 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 180, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. FENOC 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 its issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION r '-~-)tll/4tVL----

Doug1as A. Broaddus, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: Apri 1 8, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 180 RENEWED FACILITY OPERATING LICENSE NO. NPF-73 DOCKET NO. 50-412 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page 4 Page 4 Beaver Valley Power Station Units 1 and 2 share a common Appendix A, Technical Specifications. As such, the replacement pages listed in the attachment to License Amendment No. 293 will also be applicable for Amendment No. 180.

(b) Further, the licensees are also required to notify the NRC in writing prior to any change in: (i) the term or conditions of any lease agreements executed as part of these transactions; (ii) the BVPS Operating Agreement, (iii) the existing property insurance coverage for BVPS Unit 2, and (iv) any action by a lessor or others that may have adverse effect on the safe 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 set forth in 10 CFR Chapter 1 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 FENOC is authorized to operate the facility at a steady state reactor core power level of 2900 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 18 O , and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. FENOC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 18 O Beaver Valley Unit 2 Renewed Operating License NPF-73

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 293 AND 180 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-66 AND NPF-73 FIRSTENERGY NUCLEAR OPERATING COMPANY FIRSTENERGY NUCLEAR GENERATION, LLC OHIO EDISON COMPANY THE TOLEDO EDISON COMPANY BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-334 AND 50-412

1.0 INTRODUCTION

By application dated April 16, 2014 (Reference (Ref.) 1), as supplemented by letters dated November 4, 2014 (Ref. 2), and March 23, 2015 (Ref. 3), the FirstEnergy Nuclear Operating Company, et al. (the licensee), requested changes to the technical specifications (TS) for Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2).

The requested change would revise TS 5.5.12, "Containment Leakage Rate Testing Program,"

to delete reference to the BVPS-1 exemption transmittal letter dated December 5, 1984 (Ref. 4),

and require compliance with the Nuclear Energy Institute (NEI) topical report NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR

[Title 10 of the Code of Federal Regulations] Part 50, Appendix J" (Ref. 5), in lieu of Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program" (Ref. 6), including listed exceptions. Additionally, the revision of the TS would require compliance with the limitations and conditions specified in Section 4.0 of NEI 94-01, Revision 2-A (Ref. 7), consistent with an NRC letter dated August 20, 2013 (Ref. 8).

In accordance with the guidance in NEI 94-01, Revision 2-A, the proposed changes would permit the performance-based primary containment Integrated Leak Rate Testing (ILRT), also known as a Type A test, interval to be extended from no longer than 10 years to no longer than 15 years, provided acceptable performance history and other requirements stated in this report are maintained. In accordance with the guidance in NEI 94-01, Revision 3-A, the proposed changes would permit the containment isolation valve Local Leakage-Rate Test (LLRT), also known as a Type C test, interval to be extended from no longer than 60 months to no longer than 75 months, with a permissible extension period of 9 months (total of 84 months) for

non-routine emergent conditions, based on acceptable performance history as defined in NEI 94-01, Revision 3-A.

The supplements dated November 4, 2014, and March 23, 2015, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register (FR) on August 5, 2014 (79 FR 45477).

2.0 REGULATORY EVALUATION

The licensee requested a change to the Facility Operating Licensee for BVPS-1 and 2, in accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit." 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 to 10 CFR Part 50, includes two options to meet Appendix J: "Option A-Prescriptive Requirements," and "Option B-Performance-Based Requirements." The testing requirements in 10 CFR Part 50, Appendix J, Option B ensure that: (a) leakage through containments or systems and components penetrating containments does not exceed allowable leakage rates specified in the TS; and (b) integrity of the containment structure is maintained during the service life of the containment.

The regulations in 10 CFR Part 50, Appendix J Option B specify performance-based requirements and criteria for preoperational and subsequent leakage rate testing. These requirements are met by performing 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; and Type C pneumatic tests to measure containment isolation valve leakage rates. After the preoperational tests, these Type A tests must be conducted at periodic intervals based on the historical performance of the overall containment system, and Type B and C tests conducted at periodic intervals based on the safety significance and historical performance of each boundary and isolation valve to ensure 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 (La) with margin, as specified in the TS. 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.3of10 CFR 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 must be included, by general reference, in the plant TSs. 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.

NEI 94-01, Revisions 2-A and 3-A, have been reviewed by the U.S. Nuclear Regulatory Commission (NRC) and approved for use. The final safety evaluation (SE) for Revision 2-A, issued by letter dated June 25, 2008 (Ref. 9), and consistent with an NRC letter to NEI dated August 20, 2013 (Ref. 8), documents the NRC's evaluation and acceptance of Revision 2-A,

subject to six specific limitations and conditions listed in Section 4.1 of the SE for the Type A test. The final SE for Revision 3-A, issued by letter dated June 8, 2012 (Ref. 10), includes two specific limitations and conditions listed in Section 4.0 of the SE for Type C test. The licensee's submittal was reviewed against the conditions and limitations presented in the SE for Revision 2-A and Revision 3-A of NEI 94-01.

In accordance with the guidance in NEI 94-01, Revision 2-A, the licensee proposes to extend the containment Type A test interval from the current approved 10 years to 15 years, based on acceptable performance. This would allow the next Type A test to be performed within 15 years from the last test (April 15, 2006, for BVPS-1, and May 11, 2008, for BVPS-2), instead of the current 10-year interval.

In accordance with the guidance in NEI 94-01, Revision 3-A, the licensee proposes to extend the containment Type C test interval from the current approved 60 months to 75 months, with a permissible extension period of 9 months (total of 84 months) for non-routine emergent conditions, based on acceptable performance. This would allow the next Type C test to be performed within 75 months from the last test (fall 2013 for BVPS-1, and fall 2012 for BVPS-2),

instead of the current 60-month interval.

3.0 TECHNICAL EVALUATION

3.1 Licensee's Proposed Changes BVPS-1 and 2 TS 5.5.12.a, "Containment Leakage Rate Testing Program," currently states that:

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. For Unit 1, exemptions to Appendix J of 10 CFR 50 are dated November 19, 1984, December 5, 1984, and July 26, 1995. For Unit 2, exemptions to Appendix J of 10 CFR 50 are as stated in the Operating License.

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. For Unit 1, the next Type A test performed after the May 29, 1993 Type A test shall be performed no later than May 28, 2008.

2. For Unit 2, the next Type A test performed after the November 10, 1993 Type A test shall be performed no later than November 9, 2008.

A Request for Additional Information (RAI) was sent to the licensee on October 1, 2014 (Ref. 11 ), to discuss the reference used in the TSs for extending the Containment Type A and Type C leak rate testing intervals. In its response (Ref. 2), the licensee agreed to include a reference to the conditions and limitations of NEI 94-01, Revision 2-A.

In Reference 1, Attachment 1, the licensee stated BVPS TS 5.5.12.a, "Containment Leakage Rate Testing Program," would be revised to state:

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. For Unit 1, exemptions to Appendix J of 10 CFR 50 are dated November 19, 1984, and July 26, 1995. For Unit 2, exemptions to Appendix J of 10 CFR 50 are as stated in the Operating License. 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," and conditions and limitations specified in NEI 94-01, Revision 2-A.

The licensee's proposed amendment would revise the BVPS leakage rate testing program by deleting reference to the BVPS-1 exemption transmittal letter dated December 5, 1984 (Ref. 4 ),

and implementing the guidance in NEI 94-01, Revision 3-A, as well as the conditions and limitations specified in NEI 94-01, Revision 2-A. The material presented in NEI 94-01, Revision 3-A and the conditions and limitations as noted in the associated SEs for both NEI 94-01, Revisions 2-A (Ref. 7) and 3-A (Ref. 5) were used by the NRC staff to conduct the review for BVPS license amendment application.

The licensee proposed to follow NEI 94-01, Revision 3-A, and the limitations and conditions of Section 4.0 of the NEI 94-01, Revision 2-A SE, and Section 4.0 of the NEI 94-01, Revision 3-A SE. The licensee requests an extension of the Type A test interval, which is currently required by TS to be performed at 10 year intervals, to no longer than 15 years from the last Type A test (April 15, 2006, and May 11, 2008, for BVPS-1 and 2, respectively). NEI 94-01, Revision 3-A provides a guideline that an extension of the Type A test interval be based on two consecutive successful Type A tests (performance history), in addition to other guidelines stated in Section 9.2.3 in NEI 94-01, Revisions 2-A and 3-A.

The licensee also proposes an extension of the Type C test interval, from the TS requirement of 60-month intervals, to 75-month intervals, with a permissible extension period of 9 months (total of 84 months) for non-routine emergent conditions, from the last Type C test. NEI 94-01, Revision 3-A guidance explains that extensions of Type C test intervals are allowed based upon completion of two consecutive periodic as-found tests where the results of each test are within a licensee's allowable administrative limits, including other guidelines stated in Section 10.2.3 in NEI 94-01, Revision 3-A.

3.2 Deterministic Considerations: Structural and Leak-Tight Integrity of the Containment The BVPS-1 and 2 containment buildings are reinforced concrete, steel-lined vessels with a flat base, cylindrical walls, and a hemispherical dome. The foundation mat is a soil bearing concrete slab approximately 10-feet thick. The inside faces of the containment wall, dome, and mat are lined with steel liner plates. The containment buildings do not require the participation of the liner as a structural component. No credit is taken for the presence of the steel liner in designing the containment buildings to resist earthquake forces or other design loads. The liner plate serves as an impervious membrane whose function is to act as a gas tight boundary and to transmit loads to the concrete.

The cylindrical portion of the liner is 3/8-inch thick; the hemispherical dome liner is 1/2-inch thick; and the floor liner covering the mat is 1/4-inch thick. The floor liner plate is covered with a thick layer of reinforced concrete that insulates it from temperature effects.

The interior of the containment is lined with steel plates welded together to form a leak tight barrier. Since the base slab liner plate is covered with concrete, access for personnel and equipment is provided by steel penetrations that are attached to the liner plate and anchored into the concrete structure. The primary containment wall contains penetrations for different purposes, such as fuel transfer, piping, and electrical. A large diameter equipment hatch provides for transfer of material and equipment. The hatch and lock doors are supplied with compression seals with provisions for leak testing.

3.2.1 Containment lnservice Inspection (CISI) Program In Reference 1, the licensee stated that the Containment Structural Integrity Test procedures for BVPS-1 and 2 are utilized to p~rform general visual observations of the accessible interior and exterior surfaces of the containment structure in order to identify evidence of deterioration that may affect the containment structural integrity or leak tightness in accordance with:

• TS Surveillance Requirement 3.6.1.1, which requires, in part, visual examinations in accordance with the Containment Leak Rate Testing Program.

• TS 5.5.12.a, which requires, in part, visual examinations in accordance with the guidelines contained in RG 1.163 (Ref. 6). (Regulatory Position 3 states that these examinations should be conducted prior to initiating a Type A test and during two other refueling outages before the next Type A test if the interval for the Type A test has been extended to 10 years, in order to allow for early uncovering of evidence of structural deterioration.)

• American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code)Section XI, Subsections IWE and IWL require visual examinations.

General visual observations of the accessible interior and exterior surfaces of the containment structure are performed on a frequency that meets ASME Code Section XI, Subsections IWE and IWL, and 10 CFR 50 Appendix J, Option 8.

In Reference 1, the licensee also stated that the proposed change would revise TS 5.5.12 by replacing the reference to RG 1.163 with reference to NEI 94-01, Revision 3-A (Ref. 5). This will allow the inspection interval to be extended to 15 years and require that a general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leak-tight integrity be conducted prior to each Type A test and during at least three other outages before the next Type A test. These examinations are also utilized to perform inspections in accordance with ASME Code Section XI, Subsections IWE and IWL. Additionally, the tests will continue to be performed to meet the requirements of TS 5.5.12 with the incorporation of NEI 94-01, Revision 3-A guidelines and the Revision 2-A conditions and limitations.

3.2.2 Historical Type A Test Results In Section 3.1.1 of the licensee's application (Ref. 1), the licensee presented the results of the historical Type A tests, which are summarized in Table 1 and Table 2.

Table 1: BVPS-1 Type A Test Historical Results Test As Found As Found As Left Leak As Left Pressure at Completion Leak Rate Acceptance Rate Acceptance Start of Test Date* Criteria Criteria 5/29/1993 0.0150 s 0.075 0.0133 s 0.075 57.51 psia

%wt/day %wt/day %wt/day %wt/day 4/15/2006 0.0428 s 0.1 0.0350 s 0.075 59.57 psia

%wt/day %wt/day %wt/day %wt/day

• Date depressurization of containment completed

%wt/day = Percent containment air weight per day Psia =Pounds per square inch absolute As shown in Table 1, above, the Type A tests for BVPS-1 were all successfully completed.

There are no planned modifications for BVPS-1 that would require additional Type A testing, which would be due April 2016. Since the licensee has performed the Type A tests successfully, this supports the licensee's request to extend the Type A testing interval for BVPS-1 in accordance with NEI 94-01, Revision 2-A (Ref. 7), subject to the appropriate limitations and conditions discussed in sections 3.3.1 and 3.3.2 of this SE. Following approval of this LAR, the next required Type A test for BVPS-1 will be on or before April 15, 2021.

Table 2: BVPS-2 Type A Test Historical Results Test As Found As Found As Left Leak As Left Pressure at Completion Leak Rate Acceptance Rate Acceptance Start of Test Date* Criteria Criteria 11/11/1993 0.0410 s 0.075 0.0401 s 0.075 61.18 psia

%wt/day %wt/day %wt/day %wt/day 5/11/2008 0.0588 s 0.1 0.0587 s 0.075 60.65 psia

%wt/day %wt/day %wt/day %wt/day

• Date depressurization of containment completed

%wt/day = Percent containment air weight per day Psia = Pounds per square inch absolute Table 2, above, shows the results of Type A testing for BVPS-2. All Type A tests were successfully completed and support the licensee's request to extend the Type A testing interval in accordance with NEI 94-01, Revision 2-A. BVPS-2 has a steam generator replacement scheduled during the refueling outage in 2020, as stated in the March 23, 2015, supplement (Ref. 3). This change necessitates a construction opening in the reactor building and the subsequent restoration of the concrete structure. BVPS-2 plans to perform its next Type A test prior to returning the unit to service. Following the steam generator replacement necessitated containment testing, and pending approval of this LAR, the next Type A test will be performed on or before May 11, 2023, for BVPS-2.

3.2.3 Historical Type B and Type C Leak Rate Results In Section 3.1.2 of the licensee's application (Ref. 1), the licensee presented the results of the Type Band C testing. The results are summarized in Table 3 and Table 4.

Table 3: BVPS-1 Historical Type B and Type C Leak Rate Results Refueling As-Found Min Percentage of As-Left Max Percentage of Outage Path 0.6 La Path 0.6 La 1R15Spring 454.47 SCFD 11.6% 1,409.58 SCFD 35.9%

2003 1R16 Fall 2004 543.81 SCFD 13.8% 1,074.90 SCFD 27.4%

1R17 Spring 1,305.76 SCFD 33.2% 1,560.95 SCFD 37.7%

2006 1R18 Fall 2007 853.52 SCFD 20.7% 2,180.47 SCFD 52.8%

1R19Spring 581.92 SCFD 14.1% 1,330.50 SCFD 32.2%

2009 1R20 Fall 2010 858.60 SCFD 21.0% 1,563.58 SCFD 38.2%

1R21 Spring 818.46 SCFD 20.0% 1,879.23 SCFD 45.9%

2012 1R22 Fall 2013 1,007.97 SCFD 24.6% 1,680.68 SCFD 41.0%

=

SCFD Standard Cubic Feet per Day The limit for leakage is 0.6 La for both the as-found minimum pathway leak rate and the as-left maximum pathway leak rate. For BVPS-1, the average as-found minimum pathway leak rate is 19.9 percent of 0.6 La and high of 33.2 percent. Similarly, the average and high as-left maximum pathway leak rate was 38.9 percent and 52.8 percent, respectively.

Table 4: BVPS-2 Historical Type B and Type C Leak Rate Results Refueling As-Found Min Percentage of As-Left Max Percentage of Outage Path 0.6 La Path 0.6 La 2R10 Fall 2003 689.08 SCFD 16.6% 1,318.30 SCFD 31.7%

2R11 Spring 2, 706.1 SCFD 65.0% 1,374.06 SCFD 33.0%

2005 2R12 Fall 2006 966.01 SCFD 23.2% 1,173.47 SCFD 27.8%

2R13 Spring 664.99 SCFD 15.8% 1,308.06 SCFD 31.0%

2008 2R14 Fall 2009 814.81 SCFD 19.3% 1, 173.47 SCFD 27.8%

2R15 Spring 522.17 SCFD 12.4% 1,046.45 SCFD 24.8%

2011 2R16 Fall 2012 620.00 SCFD 14.7% 1,264.61 SCFD 30.0%

SCFD = Standard Cubic Feet per Day The limit for leakage is 0.6 La for both the as-found minimum pathway leak rate and the as-left maximum pathway leak rate. For BVPS-2 the average as-found minimum pathway leak rate is 23.9 percent of 0.6 La and high of 65.0 percent. Similarly, the average and high as-left maximum pathway leak rate was 31.3 percent and 41.0 percent, respectively.

3.2.4 Type Band Type C Testing Corrective Actions In the event Type Band Type C testing presented any adverse conditions during the Appendix J testing, the licensee addressed each condition in its corrective action program. Corrective actions resulted in either equipment maintenance or procedural changes. Examples of conditions include missed surveillances, such as is the case with a component cooling water supply to a reactor coolant pump penetration, and containment penetration leakage, which exceeds the limit including electrical penetrations and nitrogen supply to the pressurizer relief tank.

The missed surveillance of the BVPS-1 component cooling water supply to the reactor coolant pump penetration was due to a misinterpretation of the surveillance requirements of the work performed. There was a replacement of a relief valve and as it was new, should have been leak tested in the next refueling outage. However, there was a misinterpretation of the frequency requirements. The penetration was placed on extended frequency prior to two successful consecutive periodic as-found Type C tests. When the licensee recognized the misinterpretation, this penetration was added to the scope of testing for the current outage and an extent of condition evaluation was performed for all Type C tested components. The licensee did not find additional misinterpretations of testing frequency and this particular penetration tested successfully two consecutive times.

The licensee describes two occasions where electrical penetrations in BVPS-1 have exceeded their leakage limits. In the first, a Type Bleak test, the source of leakage could not be found and the penetration was scheduled for replacement in the following refueling outage. After the penetration replacement, the leak test was performed acceptably. In the second electrical penetration, it was determined that the source of leakage was a test connection. The test connection was sealed and then the connection re-tested successfully.

A BVPS-1 nitrogen supply to the pressurizer relief tank penetration exceeded the leakage limit.

The licensee repaired the leaking valve and re-tested, successfully. This valve was then placed on a 30-month testing frequency to be tested during the next refueling outage.

Program requirement changes were also made. In 2012, the Appendix J and Air Operated Valve (AOV) programs changed with regard to actuator packing. Previously, packing adjustments were considered non-impact, and returning and repacking was restoring the valve to its previous condition. Relaxation of packing occurs as a result of valve stroking, which may thereby cause leakage. However, there was no test data to determine if packing was tightened too much, which would increase friction. In the AOV program, pneumatic actuator packing adjustments are considered to impact Type C test limits and the program limit on total valve friction. Packing adjustments are accompanied by AOV diagnostic testing, which provides a seat load measurement that can be directly linked to an acceptable Type C test. Additionally, AOVs that are also containment isolation valves, also require Type C testing unless previous diagnostic testing provides satisfactory results.

The licensee evaluated actuator packing adjustments to determine if any Type C testing was missing. They found two penetrations in BVPS-1 and one penetration in BVPS-2 where the leak test was not performed. The three penetrations were subsequently tested and the measured leakage rates were acceptable.

3.2.5 Information Notice 92-20, "Inadequate Local Leak Rate Testing" NRC Information Notice 92-20, "Inadequate Local Leak Rate Testing," was issued to alert licensees to accuracy problems of 10 CFR 50, Appendix J local leak rate testing under accident conditions. The concern was that during testing the two plies in the bellows were in contact with each other and restricted the flow of the test medium to the crack locations. Any two-ply bellow with similar construction was susceptible to the accuracy problem.

Beaver Valley determined if any penetrations were equipped with the bellows assemblies. In BVPS-1 they found that only the recirculation spray heat exchanger metal expansion joints required Type Bleak testing. Testing is completed in accordance with the containment leak rate testing program with acceptance criteria of no visible leakage.

At BVPS-2, the licensee determined that none of the bellows assemblies were part of the reactor containment building pressure boundary and are therefore not local leak rate tested.

3.2.6 Deterministic Considerations Summary The results of the past containment leakage testing and the CISI programs demonstrate acceptable performance of the BVPS primary containment and demonstrate that the structural and leak-tight integrity of the primary containment structure is adequately managed. The structural and leak-tight integrity of the BVPS primary containment will continue to be periodically monitored and managed by the containment leakage testing and CISI programs.

Therefore, the NRC staff finds that there is reasonable assurance that the containment structural and leak-tight integrity will continue to be maintained, without undue risk to public health and safety, if the current Type A containment leakage testing program interval at BVPS is extended to 15 years.

3.3 NRC Staff Evaluation of the Conditions and Limitations RG 1.163 (Ref. 6) provides a method acceptable to the NRC for implementing the performance-based option (Option B) of 10 CFR 50, Appendix J. The regulatory positions stated in RG 1.163, as modified by NRC safety evaluation reports (SEs) (Ref. 9 and 10) are incorporated in NEI NEI 94-01, Revision 3-A (Ref. 5). Revision 2-A of NEI 94-01 (Ref. 7) included provisions for extending the Integrated Leak Rate Testing (ILRT), Type A interval, to 15 years, subject to the limitations and conditions provided in the SE for Revision 2-A, as addressed in Reference 8. Revision 3-A included guidance for extending the Type C LLRT interval to 75 months, with a permissible extension period of 9 months (total of 84 months) for non-routine emergent conditions. Type C testing ensures that individual containment isolation valves are essentially leak tight. In addition, aggregate Type C leakage rates support the leakage tightness of the primary containment by minimizing potential leakage paths.

3.3.1 NRC Conditions In NEI 94-01, Revision 2-A In the NRC SER, dated June 25, 2008 (Ref. 9), the staff concluded that the guidance in 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 addressed and satisfied these conditions, as applicable, in the LAR as discussed below.

a. NRC Condition 1 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-A, in lieu of that in ANS/ANS-56.8-2002. (Refer to SE Section 3.1.1.1 )."

The licensee states that it will utilize the definition in NEI 94-01, Revision 3-A, Section 5.0. This approach is acceptable because the definition remained unchanged from Revision 2-A to Revision 3-A of NEI 94-01. Therefore, the licensee addressed and satisfied NRC Condition 1.

b. NRC Condition 2 NRC Condition 2 states: "The 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)."

The licensee states that it will continue to perform general visual observations of the accessible interior and exterior surfaces of the containment structure, in accordance with containment structural integrity test procedures, to meet the requirements of the proposed revision to TS 5.5.12, the inspection requirements of ASME Code Section XI, subsections IWE and IWL, as outlined in a schedule in section 3.2.1 of the licensee's submittal, and NEI 94-01, Revision 3-A, Sections 9.2.1 and 9.2.3.2. The approach to use NEI 94-01, Revision 3-A is acceptable because Sections 9.2.1 and 9.2.3.2 are identical in both revisions and the licensee has submitted a schedule of inspections to be performed prior to and between Type A Tests. Therefore, the licensee addressed and satisfied NRC Condition 2.

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)."

The licensee states that it will continue to perform general visual observations of the accessible interior and exterior surfaces of the containment structure in accordance with containment structural integrity test procedures to meet the requirements of the proposed revision to TS 5.5.12, the inspection requirements of ASME Code Section XI, subsections IWE and IWL, and NEI 94-01, Revision 3-A, Sections 9.2.1 and 9.2.3.2.

The approach to use NEI 94-01, Revision 3-A is acceptable because Sections 9.2.1 and 9.2.3.2 are identical in both revisions and address containment structure areas that are potentially subject to degradation. Therefore, the licensee addressed and satisfied NRC Condition 3.

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)."

The licensee has a steam generator replacement scheduled for BVPS-2 during the refueling outage in 2020 (Ref. 3). The steam generator replacement will require a construction opening in the reactor containment. Following the replacement of the construction opening and restoration of the concrete structure, the next BVPS-2 Type A test will be performed prior to returning the unit to service. Therefore, the licensee addressed and satisfied NRC Condition 4.

e. NRC Condition 5 NRC Condition 5 states: "The normal Type A test interval should be less than 15 years. 1 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 that the previous BVPS-1 Type A test was performed on April 15, 2006 and BVPS-2 Type A test was performed on May 11, 2008. The licensee stated in its submittal that the next Type A test for BVPS-1 will be performed on or before April 15, 2021 and that BVPS-2 will be performed on or before May 11, 2023.

Extending the ILRT interval beyond 15 years does not apply to this amendment as the licensee only submitted a request for an extension up to 15 years. However, in the event that an extension beyond 15 years is required, the licensee stated it would request the extension via license amendment request in accordance with the staff position in 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" (Ref. 12). FE NOC states it will also demonstrate, at that time, that an unforeseen emergent condition exists, as specified in RIS 2008-27 and Reference 9.

The licensee has addressed and satisfied the intent of the applicable portion of NRC Condition 5 because it proposes a test interval of up to 15 years.

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, 1 Although NRG Condition 5 states that the normal Type A test interval should be less than 15 years, the NRG approved the use of Revision 2 to extend Type A test intervals up to 15 years, provided the conditions are satisfied, as described in Ref. 9 sections 4.1 and 5.0.

Revision 2, and EPRI Report No. 1009325, Revision 2, including the use of past ILRT data."

This condition is not applicable to Beaver Valley Power Station. The licensee was not licensed under 10 CFR Part 52.

3.3.2 NRC Conditions in NEI 94-01, Revision 3-A In the NRC SER, dated June 8, 2012 (Ref. 10), the staff concluded that the guidance in NEI 94-01, Revision 3-A, is acceptable for reference by licensees proposing to amend their TS in regards to containment leakage rate testing, subject to the two limitations and conditions noted as in Section 4.0 of the NRC SER for NEI 94-01, Revision 3-A. The NRC staff evaluated whether the licensee addressed and satisfied these conditions in the LAR as discussed below.

a. NRC Condition 1 NRC Condition 1 states, in part, that:

The staff is allowing the extended interval for Type C LLRTs be increased to 75 months with the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit. In addition, a corrective action plan shall be developed to restore the margin to an acceptable level. The staff is also allowing the non-routine emergent extension out to 84-months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI 94-01, Revision 3.

At no time shall an extension be allowed for Type C valves that are restricted categorically (e.g. BWR MSIVs [boiling-water reactor main steam isolation valves]), and those valves with a history of leakage, or any valves held to either a less than maximum interval or to the base refueling cycle interval.

Only non-routine emergent conditions allow an extension to 84 months.

The licensee stated that its post-outage report, as required by section 12.1 of NEI 94-01, Rev. 3-A, will include the margin between the Type B and Type C minimum pathway leak rate summation value, as adjusted, to include the estimate of applicable Type C leakage understatement, and its regulatory limit of 0.60 La.

The licensee will complete an analysis and determine the appropriate corrective action plan when the potential leakage understatement adjusted Type B and Type C minimum pathway leak rate total is greater than the BVPS administrative leakage summation limit of 0.50 La and less than the regulatory limit of 0.60 La. The corrective action plan shall focus on the components, which have contributed the most to the increase in the leakage summation value and the manner of timely corrective action, as deemed appropriate that best focuses on the prevention of future component leakage performance issues.

Consistent with the generic approval in NEI 94-01, Rev. 3-A, Beaver Valley stated it will only utilize the 9-month grace period beyond 75 months to eligible Type C components for non-routine emergent conditions, as specified in Reference 10. These occurrences will be documented in the record of tests.

The licensee has addressed and satisfied NRC Condition 1.

b. NRC Condition 2 NRC Condition 2 states in part, that:

When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B and C total, and must be included in a licensee's post-outage report.

The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

In response to NRC Condition 2, FENOC states that it will conservatively apply a potential leakage understatement adjustment factor of 1.25 to the as-left leakage total for each Type C component currently on the 75 month extended test interval. The adjustment factor of 1.25 was chosen because the change from a 60 month extended test interval to a 75 month interval is a change of 25 percent. The result is a combined conservative Type C total for all 75 month local leak rate tests being carried forward.

The adjustment factor will be included whenever the total leakage summation is required to be updated. An analysis and corrective action plan will be prepared when the summation of the potential leakage understatement adjusted leak rate total for Type C components being tested on a 75 month extended interval and the total of the Type B tested components when the summation is greater than the licensee's administrative limit of 0.50 La. but less than the regulatory limit of 0.60 La.

If the potential leakage understatement adjusted minimum pathway leak rate is less than the 0.50 La administrative leakage summation limit, then the extension for all affected Type C components has been adequately demonstrated and the calculated local leak rate total represents the actual leakage potential of the penetrations.

The licensee states, "An adverse trend is defined as three consecutive increases in the final pre-reactor coolant system Mode change Type B and Type C minimum pathway leak rate summation value adjusted to include the estimate of applicable Type C leakage understatement, as expressed in terms of La."

The licensee will develop a corrective action plan in the event that an adverse trend is observed. The plan will focus on the components that have contributed the most to the adverse trend in the leakage summation value.

Based on the review of the licensee's submittal, the NRC staff concludes that the primary containment leakage rate testing program contains provisions for trending and monitoring that conservatively applies a leakage understatement factor to account for the extended interval. In addition, the post-outage report will contain the necessary information. Therefore, the licensee addressed and satisfied NRC Condition 2.

3.4 Probabilistic Risk Assessment (PRA) 3.4.1 Background NEI 94-01, Revision 3-A (Ref. 5), Section 9.2.3.4, "Plant-Specific Confirmatory Analyses,"

states, in part, that the assessment should be performed using the approach and methodology described in Electric Power Research Institute (EPRI) Technical Report (TR) 1009325, Revision 2-A (Ref. 7), "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.

In the SER dated June 25, 2008 (Ref. 9), the NRC staff found the methodology in EPRI TR-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 NRC SER for EPRI TR-1009325, Revision 2 (Ref. 13), stipulate that:

1. The licensee submit documentation indicating that the technical adequacy of their PRA is consistent with the requirements of RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" (Ref. 14 and 15), relevant to the ILRT extension application.

2. The licensee submits documentation indicating that the estimated risk increase associated with permanently extending the ILRT surveillance interval to 15 years is small and consistent with the clarification provided in Section 3.2.4.6 of the SER for EPRI TR-1009325, Revision 2.

3. The methodology in EPRI TR-1009325, Revision 2, is acceptable provided 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. An LAR is required in instances where containment over-pressure is relied upon for emergency core cooling system (ECCS) performance.

3.4.2 Plant-Specific Risk Evaluation The licensee performed a risk impact assessment for extending the Type A containment ILRT interval from 10 years to 15 years. The risk analyses for BVPS-1 and 2 were provided in Attachments 3 and 4 of the LAR dated April 16, 2014 (Ref. 1), respectively. Additional information was provided by the licensee in its supplement dated November 4, 2014 (Ref. 2), in response to NRC RAls.

In Section 1.1 of Attachments 3 and 4 to the LAR, the licensee stated that the plant-specific risk assessments for both units follow the guidance in:

• NEI 94-01, Revision 3-A,

• The methodology described in EPRI TR-1009325, Revision 2-A,

• The methodology used in EPRI TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals," dated August 1994,

• The NEI "Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals" from November 2001,

• The NRC regulatory guidance on the use of PRA, as stated in RG 1.200, as applied to ILRT interval extensions and risk insights in support of a request for a plant's licensing basis as outlined in RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (Ref. 16),

and

• The methodology used for Calvert Cliffs Nuclear Plant (CCNP) to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval.

The licensee stated in the LAR that BVPS-1 and 2 have Level 2 PRA models that include both internal and external events including internal floods, internal fires, and seismic events.

The licensee addressed each of the four conditions for the use of EPRI TR-1009325, Revision 2, which are listed in Section 4.2 of the NRC SER. A summary of how each condition has been met is provided in the following sections.

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

In Regulatory Issue Summary 2007-06, "Regulatory Guide 1.200 Implementation" (Ref. 17), the NRC clarified that for all risk-informed applications received after December 2007, the NRC staff will use Revision 1 of RG 1.200 (Ref. 13) to assess technical adequacy of the PRA used to support risk-informed applications. Revision 2 of RG 1.200 (Ref. 14) will be used for all risk-informed applications received after March 2010. In Section 3.2.4.1 of the SER to EPRI TR-1009325, Revision 2, the NRC staff stated, in part, that:

[l)icensee requests for a permanent extension of the ILRT surveillance interval to 15 years pursuant to NEI TR 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, will be treated by NRC staff as risk-informed license amendment requests. Consistent with information provided to industry in Regulatory Issue Summary 2007-06, "Regulatory Guide 1.200 Implementation," the NRC staff will expect the licensee's supporting Level 1/LERF PRA to address the technical adequacy requirements of RG 1.200, Revision 1.... Any identified deficiencies in addressing this standard shall be assessed further in order to determine any impacts on any proposed decreases to surveillance frequencies. If further revisions to RG 1.200 are issued which endorse additional standards, the NRC staff will evaluate any application referencing NEI TR 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, to examine if it meets the PRA quality guidance per the RG 1.200 implementation schedule identified by the NRC staff.

In the same section of the SER, the NRC staff states that Capability Category I of ASME PRA standard shall be applied as the standard for assessing PRA quality for IRL T extension applications, as 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.

Per Section 2 of Attachments 3 and 4, the confirmatory analysis for each Unit uses results from a Level 2 analysis of core damage scenarios from the current PRA models for BVPS-1 and 2 (referred to as PRA-BV1-AL-R05a and PRA-BV2-AL-R05a, respectively) and subsequent containment responses resulting in various fission product release categories. These PRA models include both internal and external events (seismic and internal fire), and provide Level 1 and Level 2 results.

The PRA technical adequacy for BVPS-1 and 2 is discussed in Section 3.5.2 of the LAR and Attachments 5 and 6 to the LAR. The licensee stated that an independent PRA peer review of the BVPS PRA models was conducted under the auspices of the Westinghouse Owners Group (WOG) in July 2002. Following PRA Model revisions in 2006, a self-assessment of the BVPS PRA models was conducted in 2007 to determine if there were any gaps present between the BVPS PRA models and meeting the Capability Category II Supporting Requirements (SR) in the 2005 version of the ASME PRA Standard Addendum B and qualifications provided in the NRC endorsement of the standard contained in RG 1.200, Revision 1. As part of the resolution to several Facts and Observations (F&Os) from the 2002 PRA peer review, a change in the Human Reliability Analysis (HRA) methodology was incorporated into the 2006 BVPS-1 PRA model revision, so a focused scope peer review of the HRA Technical Elements against the ASME PRA Standard was performed using RG 1.200, Revision 1. Finally, a focused-scope peer review of the Internal Flood PRA Technical Elements was performed for each unit against the ASME/ANS PRA standard (along with the NRC clarifications provided in RG 1.200, Revision 2) following upgrades of the internal flooding models in 2010.

In Table 2 of Attachments 5 and 6, the licensee provided a brief summary of the BVPS final resolutions to all of the 2007 BVPS PRA Self-Assessment, 2007 BVPS HRA Focused Peer Review, and the 2011 BVPS Internal Flood PRA Focused Peer Review F&Os, which resulted in a change to the PRA model. The licensee further stated that "[a]ll other F&Os from these assessment/reviews were considered to be documentation issues, and did not impact the PRA models."

Given that the implementation date of RG 1.200, Revision 2, was April 2010 and the BVPS internal events PRA models were reviewed against the ASME RA-Sb-2005 PRA standard, the NRC staff requested that the licensee identify any gaps between the BVPS PRA models used in this application and RG 1.200, Revision 2 requirements. In its response to PRA Licensing Branch (APLA) RAI 1 (Ref. 2), the licensee stated that a gap assessment was performed using Section 3.3 of NEI 05-04, "Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA standard," Revisions 3, and provided the results of the gap assessment. In response to APLA RAI 2, the licensee stated that an analysis of the adequacy of the internal events PRA models for BVPS-1 and 2 submitted to the NRC on February 14, 2014 (Ref. 18) included final resolution of F&Os that did not meet ASME/ANS PRA Standard Capability Category I requirements.

In Section 3.2.4.2 of the SER for NEI 94-01, Revision 2 and EPRI TR-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."

In Sections 3.0 and 5.0 of Attachments 4 and 6, the licensee stated that BVPS-1 and 2 seismic and internal fire PRA models have not been assessed by a PRA Peer Review against the requirements of the ASME/ANS PRA Standard. The licensee further stated that to the extent that seismic and internal fire accident sequence logic is incorporated into the internal events PRA system event tree logic, they have had some limited peer checks. In response to APLA RAI 3 to discuss whether using the peer-reviewed Fire PRA model would significantly change the total estimated LERF, conditional containment failure probability (CCFP) and increase in the total population dose, the licensee stated that the modifications credited in the peer-reviewed fire PRA models developed in support of adopting the National Fire Protection Association Standard 805 are not yet installed and therefore the models do not represent the as-built, as-operated plants. The licensee further stated that the current PRA models used for this application modeled seismic and internal fire using detailed full-scope Level 2 PRA models, are based on IPEEE methodology, and have been subjected to independent review during the IPEEE submittal evaluation process.

The NRC staff reviewed Table 2 of Attachments 5 and 6 of the LAR and the licensee's responses to APLA RAI 1 and APLA RAI 2 to determine the technical adequacy of BVPS internal events models for this application. Given that the licensee evaluated its PRA against Revision 1 of RG 1.200 and the ASME PRA standard, performed a gap assessment of internal events PRA against Revision 2 of RG 1.200, and evaluated the findings developed during the reviews of its PRA for applicability to the ILRT extension, 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.

3.4.2.2 Estimated Risk Increase The second condition stipulates that the licensee submit documentation indicating that the estimated risk increase associated with permanently extending the ILRT interval to 15 years is small, and consistent with the guidance in RG 1.174 and the clarification provided in Section 3.2.4.6 of the NRC SER for NEI 94-01, Revision 2-A (Ref. 9). 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 one percent of the total population dose, whichever is

less restrictive. In addition, a small increase in CCFP should be defined as a value marginally greater than that accepted in previous one-time 15-year ILRT extension requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage points. Additionally, for plants, such as BVPS-1 and 2, that rely on containment over-pressure for net positive suction head (NPSH) for ECCS injection, both CDF and LERF should be considered in the ILRT evaluation and compared with the risk acceptance guidelines in RG 1.174. Thus, the associated risk metrics include: CDF, LERF, population dose, and CCFP.

The licensee reported the results of the plant-specific risk assessment in Section 3.5.3 of the LAR. Details of the risk assessment for BVPS-1 and 2 are provided in Attachments 3 and 4 of the LAR, respectively. The reported risk impacts are risk impact from baseline, which estimates the impact of a change in test frequency from three tests in 10 years (the test frequency under 10 CFR 50 Appendix J, Option A) to one test in 15 years and risk impact from current, which estimates the impact of a change from one test in 10 years to one test in 15 years. The following conclusions can be drawn based on the licensee's analysis associated with extending the Type A ILRT frequency:

1. The reported increase in total (internal and external) CDF for a change in test frequency from three tests in 10 years to one test in 15 years is 3.58 x 10-7 per year for BVPS-1 and 7

6.02 x 10* for BVPS-2 (Table 6-5 of Attachments 3 and 4). These numbers include the corrosion impact. These changes in CDF are considered to be "very small" (i.e., below 1 x 10*5 per year) per the acceptance guidelines in RG 1.174.

2. The reported increase in LERF for a change in test frequency from three tests in 10 years to one test in 15 years is 2.15 x 10-7 per year for BVPS-1 and 1. 63 x 10-7 per year for BVPS-2 (from Table 6-5 of Attachments 3 and 4 of the LAR). These results include both internal and external events (internal fires and seismic events) and the impacts from corrosion and loss of containment overpressure due to a large pre-existing containment liner leak. These changes in internal and external events risk are considered to be "small" (i.e., between 1 x 10"6 per year and 1 x 10*7 per year) per acceptance guidelines in RG 1.174. According to RG 1.174, use of the "small" acceptance guideline requires an assessment of baseline LERF to show that the total LERF is less than 1 x 10*5 per reactor year. In Attachments 5 and 6 of the LAR, the licensee estimated the total base LERF to be 9.64 x 10-s per year for BVPS-1 and 7

2.18 x 10* per year for BVPS-2. Thus, the new total LERFs, given the increase in ILRT interval, would be approximately 3.1 x 10*7 per year for BVPS-1 and 3.8 x 10-7 per year for BVPS-2, which are below the total LERF value of 1.0 x 10-5 per reactor year in RG 1.174.

In Section 6.1 of Attachments 3 and 4, the licensee presented sensitivity analyses to gain an understanding of the sensitivity of the results to key parameters in the corrosion analysis. Because two of the five industry corrosion events occurred at BVPS-1, a special BVPS-1 upper bound sensitivity case was performed assuming a probability of 1.0 that another hole will occur in the cylindrical steel liner during the 15 year ILRT extension. The NRC staff finds that the special upper bound sensitivity analysis for BVPS-1 is warranted and concludes that the total LERF criterion of RG 1.174 is met on the sensitivity analysis.

3. Given a change in Type A ILRT frequency from three in 1O years to one in 15 years and assuming the loss of containment overpressure at 100 La. the reported increase in the total population dose for BVPS-1 is 7 .19 x 10-1 person-rem per year, or 1.15 percent of the total population dose (Table 6-5 of Attachment 3) and for BVPS-2 is 1.21 person-rem per year, or 3.03 percent of the total population dose (Table 6-5 of Attachment 4). The reported increase in the total population dose for BVPS-2 in Table 6-5 of Attachment 4 is higher than the guideline values associated with a small increase in population dose.

In response to APLA RAI 4 (Ref. 11 ), the licensee presented an evaluation using the modular accident analysis program (MAAP) to determine the impact of operation of the most limiting recirculation spray pumps under accident conditions assuming through-wall holes in BVPS containments. The licensee concluded that "since the containment leakage rate must be greater than 1000 La in order for BVPS-1 and BVPS-2 to lose NPSH for those systems that require containment overpressure, the original analysis based on 100 La performed in Section 6.3 and Tables 6-4 and 6-5 of Attachments 3 and 4 to the LAR are considered to be overly conservative." In response to APLA RAI 4, the licensee, in part, used EPRI expert elicitation results to update its assessment assuming loss of containment overpressure at 1000 La. The updated analysis indicated a decrease in the total population dose by more than an order of magnitude. As discussed in the SER for EPRI TR-1009325, the NRC staff has not accepted the EPRI expert elicitation, as presented in the appendices of EPRI Report No. 1009325, Revision 2 and, therefore, the NRC staff does not endorse the use of expert elicitation results as presented in the licensee's updated assessment.

However, the NRC staff finds that the analysis presented in the original LAR, which assumes that the EPRI Class 3b (100 La) contribution would lead to loss of containment overpressure and make the recirculation spray (RS) pumps unavailable for such an isolation failure, was conservative as shown by the licensee's MAAP analysis. Using the EPRI TR-1009325 Revision 2-A guidance and assuming the availability of NPSH for RS pumps at 100 La, as demonstrated by the licensee, reduces the total population dose to values that are below the values associated with a small increase in population dose, as provided in Section 3.2.4.6 of the NRC SER for NEI 94-01, Revision 2. Thus, this increase in the dose for the proposed change is small and below the acceptance guidelines in Section 3.2.4.6 of the NRC SER for NEI 94-01, Revision 2.

4. The increase in CCFP due to change in test frequency from three in 10 years to one in 15 years is 1.10 percent for BVPS-1 and 1.20 percent for BVPS-2. These values are very small and are below the acceptance guidelines in Section 3.2.4.6 of the NRC SER for NEI 94-01, Revision 2.

Based on the review of the BVPS risk assessment results, the NRC staff concludes that, for BVPS-1 and 2, the increases in CDF and LERF are small and consistent with the risk acceptance guidelines of RG 1.174.

In addition, the increase in the total integrated plant risk and the small magnitude of the change in the CCFP for the requested change are small and supportive of the LAR. The defense-in-depth philosophy is maintained because the independence of barriers will not be degraded as a result of the requested change, and the use of 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.

3.4.2.3 Leak Rate for the Large Pre-Existing Containment Leak Rate Case The third condition stipulates that in order to make the methodology in EPRI TR-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 Section 3.5.1 of the LAR, the methodology in EPRI TR-1009325, Revision 2-A, incorporates the use of 100 La as the average leak rate for the pre-existing containment large leak rate accident case, and this value is used in the BVPS-1 and BVPS-2 plant-specific risk assessments. Accordingly, the third condition is met.

3.4.2.4 Applicability of Containment Over-Pressure is Credited for ECCS Performance.

The fourth condition stipulates that in instances where containment over-pressure is relied upon for ECCS performance, an LAR is to be submitted. In Section 3.5.1 of the LAR, the licensee stated that the mitigation of design basis accidents rely on containment overpressure in the calculation of available NPSH for the recirculation spray pumps at both Units and low head safety injection pumps at BVPS-1 when taking suction from the containment sump during the safety injection recirculation phase. According to the clarification provided in Section 3.2.4.6 of the NRC SER for NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2, plants that rely on containment over-pressure NPSH for ECCS injection must also consider CDF in the ILRT evaluation. The results of these analyses are included in Sections 6.3 of Attachment 3 and 4 of the LAR and in the licensee's response to APLA RAI 4. The licensee's assessment in response to APLA RAI 4 was discussed in Section 3.2.2 of this report. The increase in CDF reported in the original LAR was determined to meet the guidelines in RG 1.174 as discussed in Section 3.2.2 of the report. Accordingly, the fourth condition is met.

3.4.3 PRA Conclusion Based on the evaluation in sections 3.4.2.1 through 3.4.2.4, 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.

3.5 Evaluation of Proposed Changes to TS 5.5.12.a Deletion of Reference to Exemption The licensee proposed to remove the reference to a BVPS-1 exemption, transmitted via NRC letter dated December 5, 1984 (Ref. 4), from TS 5.5.12.a. The exemption was granted from 10 CFR 50, Appendix J, Section 111.D.1.(a), which states:

After the preoperational leakage rate tests, a set of three Type A tests shall be performed, at approximately equal intervals during each 10-year service period. The third test of each set shall be conducted when the plant is shutdown for the 10-year plant inservice inspections.

This exemption granted performance of the BVPS-1 Type A test at a 40 plus or minus 10 month frequency if the licensee were to follow the regulations under 10 CFR 50 Appendix J, Option A.

However, the current and proposed TS require that the licensee meet requirements in 10 CFR 50 Appendix J, Option B. Therefore, the Option A exemption is no longer necessary.

The NRC staff concludes that removal of the reference to this exemption in TS 5.5.12.a is acceptable.

Removal of RG 1. 163 and Addition of NE/ 94-01 Rev. 2-A and Rev. 3-A The licensee proposed to remove the following from the TS:

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995 ...

In its place, the licensee proposed the addition of the following:

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," and conditions and limitations specified in NEI 94-01, Revision 2-A.

Based on the staff's evaluation in this SE, the licensee has demonstrated that the proposed test intervals are acceptable and met applicable conditions and limitations in NEI 94-01, Rev. 2-A, as classified by the NRC staff's August 2013 letter (Ref. 8), as well as the prescribed guidelines in Rev. 3-A. Therefore, the NRC staff concludes that the above change is acceptable.

Removal of Type A Test Oates The licensee proposed to delete the following:

1. For Unit 1, the next Type A test performed after the May 29, 1993 Type A test shall be performed no later than May 28, 2008.

2. For Unit 2, the next Type A test performed after the November 10, 1993 Type A test shall be performed no later than November 9, 2008.

As these dates occur in the past, the NRC staff finds the removal of items 1 and 2 is acceptable.

3.6 Technical Evaluation Conclusion Based on the discussion in SE Sections 3.1 through 3.5, the NRC staff concludes that the proposed changes to TS 5.5.12.a are acceptable.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located with the restricted area as defined in 10 CFR Part 20. The NRC 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 propose finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (79 FR 45477, August 5, 2014). 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.

6.0 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) there is reasonable assurance that 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.

7.0 REFERENCES

1. FirstEnergy Nuclear Operating Company, "License Amendment Request to Extend Containment Leakage Test Frequency," Docket Nos. 50-334 and 50-412, April 16, 2014, Agencywide Documents Access and Management System (ADAMS) Accession No.

ML14111A291.

2. FirstEnergy Nuclear Operating Company, "Response to Request for Additional Information regarding the Technical Specification Change to Extend Containment Leak Rate Test Frequency," Docket Nos. 50-334 and 50-412, November 4, 2014, ADAMS Accession No.

ML14308A196.

3. FirstEnergy Nuclear Operating Company, "Correction of License Amendment Request to Extend Containment Leakage Rate Test Frequency (TAC Nos. MF3985 and MF3986),"

Docket Nos. 50-334 and 50-412, March 23, 2015, ADAMS Accession No. ML15082A422.

4. U.S. Nuclear Regulatory Commission, "Exemption to 10 CFR 50, Appendix J, Section 111.D.1.(a) Licensing Action TAC 55312," December 5, 1984, 49 FR 48117 (December, 10, 1984), ADAMS Accession No. ML003766713.

5. Nuclear Energy Institute, Topical Report 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.

6. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," September 1995, ADAMS Accession No. ML003740058.

7. 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.

8. U.S. Nuclear Regulatory Commission, 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,'" August 20, 2013, ADAMS Accession No. ML13192A394.

9. U.S. Nuclear Regulatory Commission, 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 Test Intervals,"'

June 25, 2008 ADAMS Accession No. ML081140105.

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ME2164}," June 8, 2012, ADAMS Accession No. ML121030286.

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Request for Additional Information RE: License Amendment Request to Extend Containment Leakage Rate Test Frequency (TAC Nos. MF3985 and MF3986}," October 1, 2014, ADAMS Accession No. ML14259A448.

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ML072970208.

14. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," January 2007, ADAMS Accession No. ML070240001.

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17. U.S. Nuclear Regulatory Commission, Regulatory Issue Summary 2007-06, "Regulatory Guide 1.200 Implementation, March 22, 2007, ADAMS Accession No. ML070650428.

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MF3301, MF3302," Docket Nos. 50-334 and 50-412, February 14, 2014, ADAMS Accession No. ML14051A499.

Principal Contributors: Taylor Lamb Diana Woodyatt Dan Hoang Mehdi Reisi Fard Date: April 8, 2015

ML15078A058 OFFICE LPL 1-2/PM LPL 1-2/LA APLA/BC SCVB/BC EMCB/BC NAME TLamb ABaxter

• SDinsmore (A) NKaripineni (A) TLupold DATE 3/24/2015 03/20/15 3/24/15 3/26/15 3/26/15 OFFICE STSB/BC OGC LPL 1-2BC LPL 1-2/PM NAME RElliott MYoung DBroaddus TLamb DATE 3/25/15 4/7/2015 4/8/2015 4/8/2015