ML14148A235

Issuance of Amendment to Extend the Containment Type a and Type C Leak Rate Test Frequency

ML14148A235
Person / Time
Site:	Surry
Issue date:	07/03/2014
From:	Shawn Williams Plant Licensing Branch II
To:	Heacock D Virginia Electric & Power Co (VEPCO)
Williams S
References
TAC MF2612, TAC MF2613
Download: ML14148A235 (39)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 3, 2014 Mr. David A. Heacock President and Chief Nuclear Officer Virginia Electric and Power Company lnnsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

SURRY POWER STATION, UNITS 1 AND 2- ISSUANCE OF AMENDMENT REGARDING THE CONTAINMENT TYPE A AND TYPE C LEAK RATE TESTS (TAC NOS. MF2612 AND MF2613)

Dear Mr. Heacock:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 282 to Renewed Facility Operating License DPR-32 and Amendment No. 282 to Renewed Facility Operating License No. DPR-37 for the Surry Power Station, Unit Nos. 1 and 2, respectively. The amendments revise Technical Specification (TS) 4.4.B, "Containment Leakage Rate Testing Requirements," in response to your application dated August 12, 2013, as supplemented by letters dated January 24, March 13, and March 25, 2014.

Specifically, the amendment extends the Type A primary containment Integrated Leak Rate Test intervals to fifteen years, extends the Type C local leak rate test intervals to 75 months, and incorporates the regulatory positions stated in RG 1.163, "Performance-Based Containment Leak-Rate Testing Program."

A copy of the Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Shawn Williams, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-280 and 50-281

Enclosures:

1. Amendment No. 282 to DPR-32

2. Amendment No. 282 to DPR-37

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 282 Renewed License No. DPR-32

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

A. The application for amendment by Virginia Electric and Power Company (the licensee) dated August 12, 2013, as supplemented by letters dated January 24, March 13, and March 25, 2014, 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; 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.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Renewed Facility Operating License No. DPR-32 is hereby amended to read as follows:

(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 282, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. DPR~32 and the Technical Specifications Date of Issuance: July 3, 2014

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO.2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 282 Renewed License No. DPR-37

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

A. The application for amendment by Virginia Electric and Power Company (the licensee) dated August 12, 2013, as supplemented by letters dated January 24, March 13, and March 25, 2014, 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; 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.

Enclosure 2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Renewed Facility Operating License No. DPR-37 is hereby amended to read as follows:

(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 282, are hereby incorporated in the renewed license. The licensee shall operate ttie facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes License No. DPR-37 and the Technical Specifications Date of Issuance: July 3, 2014

ATTACHMENT TO LICENSE AMENDMENT NO. 282 RENEWED FACILITY OPERATING LICENSE NO. DPR-32 DOCKET NO. 50-280 AND LICENSE AMENDMENT NO. 282 RENEWED FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NO. 50-281 Replace the following pages of the Renewed Facility Operating Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages License License License No. DPR-32, page 3 License No. DPR-32, page 3 License No. DPR-37, page 3 License No. DPR-37, page 3 TS TS TS page 4.4-1 TS page 4.4-1

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part40, Sections 50.54 and 50.59 of

( 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2587 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No .. 282are hereby incorporated in the renewed license. The licensee 1 .

shall operate the facility in accordance with the.Technical Specifications.

C. Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.

• D. Records .

• The licensee shall keep facility operating records in accordance with the requirements of the Technical Specifications.

E. Deleted by Amendment 65 .

F. Deleted by Amendment 71 G. Deleted by Amendment 227 H. Deleted by Amendment 227 I. Fire Protection The licensee shall implement and mail~tain in effect the provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report and as approved in the SER dated September 19, 1979, (and Supplements dated May 29, 1980, October 9, 1980, December 18, 1980, February 13, 1981, December 4, 1981, April27, 1982, November 18; 1982, January 17, 1984, February 25, 1988, and Surry- Unit 1 Renewed License No. DPR-32 Amendment No. 282

E. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. *

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

A. Maximum Power Level

• The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2587 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 282are hereby incorporated in this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

-I.

'J.:

C. Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.

D. Records The licensee shall keep facility operating records in accordance with the requirements of the Technical Specifications.

E. Deleted by Amendment 54 F. Deleted by Amendment 59 and Amendment 65 G. Deleted by Amendment 227 H. Deleted by Amendment 227

. **;4>

Surry - Unit 2 Renewed License No. DPR-37 Amendment No*. 282

TS 4.4-1 4.4 CONTAINMENT TESTS Applicability Applies to containment leakage testing.

Objective To assure that leakage of the primary reactor containment and associated systems is held within allowable leakage rate limits; and to assure that periodic surveillan-ce is performed to assure proper maintenance and leak repair during the service life of the containment.

Specification A. Periodic and post-operational integrated leakage rate tests of the containment shall be performed in accordance with the requirements of 10 CFR 50, Appendix J, "Reactor Containment Leakage Testing for Water Cooled Power Reactors."

B. Containment Leakage Rate Testing Requirements I. The containment and containment penetrations leakage rate shall be demonstrated by performing leakage rate testing as required by 10 CFR 50 Appendix J, Option B, as modified by approved exemptions. and in accordance with the guidelines contained in NEI 94-01, Revision 3-A, Industry Guidelines for Implem,enting Performance-Based Option of 10 CFR 50, Appendix J," dated July 2012.

2. Leakage rate acceptance criteria are as follows:

a. An overall integrated leakage rate of Jess than or equal to La, 0.1 percent by weight of containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, at calculated peak pressur,e (Pa).

b. A combined leakage rate of Jess than or equ~l to 0.60 La for all penetrations and valves subject to Type Band C testing when pressurized to Pa.

Prior to entering an operating condition where containment integrity is required the as-left Type A leakage rate shall not exceed 0.75 La and the combined leakage rate of all penetrations subject to Type B and C testing shall not exceed 0.6 La.

3. The provisions of Specification 4.0.2 are not applicable.

Basis The leak tightness testing of all liner welds was performed during construction by welding a structural steel test channel over each weld seam and performing soap bubble and halogen leak tests.

Amendment Nos. 282 and 282

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 282 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-32 AND AMENDMENT NO. 282 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-280 AND 50-281

1.0 INTRODUCTION

In a letter dated August 12, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13232A042), as supplemented by letters dated January 24, March 13, and March 25, 2014 (ADAMS Accession Nos. ML14035A219, ML14079A082, ML14091A262, respectively}, Virginia Electric and Power Company (Dominion) requested an amendment to Operating License DPR-32 and DPR-37 in the form of changes to the Technical Specifications {TSs) for the Surry Power Station, Units 1 and 2.

The License Amendment Request (LAR) proposes a change to Surry TS 4.4.B, "Containment Leakage Rate Testing Requirements," by replacing the reference to Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Rate Testing Program," with a reference to Nuclear Energy Institute (NEI) topical report NEI 94-01, Revision 3-A, in order to develop the Surry performance-based leakage testing program in accordance with Option B of Title 10 of the Code of Federal Regulations {1 0 CFR) Part 50, Appendix J. The purpose is to extend the Type A primary containment Integrated Leak Rate Test (ILRT) intervals to fifteen years and the Type C local leak rate test intervals to 75 months, and incorporate the regulatory positions stated in RG 1.163.

2.0 REGULATORY EVALUATION

Type A tests (also referred to as the ILRT} are tests intended to measure the primary reactor containment overall integrated leakage rate after the containment has been completed and is ready for operations and at periodic intervals thereafter. After the preoperational test, the ILRT is conducted at a periodic interval based on historical performance of the overall primary reactor containment system. Type B tests are tests intended to detect local leaks and to measure leakage across each pressure-containing or leakage-limiting boundary for the primary reactor Enclosure 3

containment penetrations. Type C tests are tests intended to measure containment isolation valve leakage rates. After the preoperational tests, Type B and Type C tests *are required to be conducted prior to initial criticality, and periodically thereafter at intervals based on the safety significance and historical performance of each bounpary and isolation valve to ensure integrity of the overall containment system and as a barrier to fission product release from reactor accidents.

10 CFR, Part 50, Appendix J, Option B, "Performance Based Requirements," requires that a Type A test be conducted at a periodic interval based on historical performance of the overall containment system. Surry TS 4.4.B, "Containment Leakage Rate Testing Requirements,"

requires that leakage rate testing be performed as required by 10 CFR Part 50, Appendix J, Option B, as modified by approved exemptions, and in accordance with the guidelines contained in U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Rate Testing Program," dated September 1995 (Accession No. ML003740058). This RG endorses, with certain exemptions, NEI report 94-01, Revision 0, "Industry Guideline for Implementing Performance Based Option of 10 CFR Part 50, Appendix J," dated July 26, 1995. '

10 CFR 50.55a "Codes and Standards," contains the containment lnservice Inspection (CISI) requirements that, in conjunction with the requirements Appendix J, ensure the continued leak-tight and structural integrity of the conta!nment during its service life.

• 10 CFR 50.65 (a), "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," states in part that the licensee shall monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components, as defined in paragraph (b) of this section, are capable of fulfilling their intended functions. These goals shall be established commensurate with safety and, where practical, take into account industry-wide operating experience.

A Type A test is an overaiiiLRT 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, but this "should be used only in cases where refueling schedules have been changed to accommodate other factors." The most recent two Type A tests at Surry have been successful, so the current interval requirement is 10 years.

Guidance for extending Type A ILRT surveillance intervals beyond ten years is provided in Nuclear Energy Institute Topical Report NEI 94-01, Revision 2, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," October 2008.

(ADAMS Accession No. ML072970206). Guidance for extending Type C Local Leak Rate Test (LLRT) surveillance intervals beyond sixty months is given in Nuclear Energy Institute Topical Report NEI94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," July 2012. (ADAMS Accession No. ML12221A202).

The Type A, Type B, and Type C test results must not exceed the La with margin, as specified in the TSs or associated bases. 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.

Dominion proposes to extend the interval for the primary containment ILRT to no longer than 15 years from the last ILRT. The last ILRT was completed on May 2006, for Unit 1 and October 2005, for Unit 2. The ILRT is currently required to be performed at a ten years and 10 months interval and is due no later than May 6, 2016, for Unit 1 and October 26, 2015, for Unit 2, as required by Surry TS 4.4.8. Using the proposed interval of no longer than 15 years, the next ILRT will be due no later than May 6, 2016, for Unit 1 and October 26, 2015 for Unit2.

As described in NRC's August 29, 2013, 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" (ADAMS Accession No. ML13192A394), Revision 3-A, inadvertently did not include the six limitations and conditions in NRC's June 25, 2008, final Safety Evaluation (SE) approving NEI 94-01, Revision 2. Although the six limitations and conditions were not included in NEI 94-01, Revision 3-A, they apply to a licensee's request to use NEI 94-01, Revision 3-A, requesting to extend the ILRT.

• NEI TR 94-01, Revision 3-A, added guidance for extending the Type C LLRT interv£1 to 75 months. Surry proposes to extend testing of Type C components from a maximum interval of 60 months to 75 months. RG 1.163, "Performance-Based Containment Leak-Test Program,"

dated September 1995 is still referenced in the new proposed TSs.

NEI TR 94-01, Revision 3-A (page iv, Executive Summary) states that: "Intervals may be increased from 30 months ... up to a maximum of 75 months for Type C tests ... If a licensee considers extended test intervals of greater than 60 months for ... Type C tested components, the review should include the additional considerations of as-found tests, schedule and review ... If the Type C test results are not acceptable, the test frequency should be set at the initial test interval~ ..

Once the cause determination and corrective actions have been completed, acceptable

• performance may be reestablished and the testing frequency returned to the extended intervals .... "

NEI TR 94-01, Revision 3-A, Section 10.2.3.3 (Type C testing) stipulate that the performance of these shall be performed at a frequency of at least once per 30 months if a penetration is replaced or engineering judgment determines that modification of a penetration has invalidated the valve's performance history; and that testing shall continue at this frequency until an adequate performance history is established ..

NEI TR 94-01, Revision 3-A, Section 10.1, states that the: "intervals of up to 60 months for the recommended surveillance frequency for ... Type C testing given in this section may be extended by up to 25 percent of the test interval, not to exceed nine months." The NRC staff agrees with this extension as being consistent with scheduling practices forTS.

10 CFR 50, Appendix J, Option B,Section V.B.3, requires that the RG 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 forTS revisions must contain justification, including supporting analyses, if the licensee chooses to deviate from methods approved by the Commission and endorsed in a.RG.

3.0 TECHNICAL EVALUATION

3.1 Containment Review 3.1 .1 Licensee's Proposed Changes In the LAR, the licensee stated that Surry TS 4.4.8, "Containment Leakage Rate Testing Program"'

currently states:

The containment and containment penetrations leakage rate shall be demonstrated by performing leakage rate testing as required by 10 CFR 50 Appendix J, Option 8, as modified by approved exemptions, and in accordance with the guidelines contained in RG 1.163, dated September, 1995 as modified by the following exception: NEI 94-01-1995, Section 9.2.3: The first Unit 2 Type A test performed after the October 26, 2000, Type A test shall be performed no later than October 26, 2015.

• The proposed amendment would revise Surry TS 4.4.8, "Containment Leakage Rate Testing

. Program," to state, The containment and containment penetrations leakage rate shall be demonstrated by performing leakage rate testing as required by 10 CFR 50 Appendix J, Option 8, as modified by approved exemptions, and in accordance with the guidelines contained in NEI 94-01, Revision 3-A, "Industry Guidelines for Implementing Performance-Based Option of 10 CFR 50, Appendix J," dated July 2012.

The proposed change in the LAR would revise the aforementioned portion of TS 4.4.8 by replacing the reference to RG 1.163 with a reference to NEI 94-01, Revision 3-A, as the implementation document. The licensee justified the proposed change by demonstrating adequate performance of the Surry containment based on historical 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 3-A. The NRC staff reviewed this LAR and supplemental information submittals from the point of deterministic considerations with regard to containment leak-tight integrity if the current Type A ILRT interval is extended from 10 years to 15 years and if the Type C LLRT test interval is extended from 60 months to 75 months. **

3.1.2 Staff's Evaluation 3.1.2.1 Integrated Leakag~ Rate Test, Type A Test Frequencies As required by 10 CFR 50.54(o), the Surry containment is subject to the requirements. set forth in 10 CFR 50, Appendix J. Option B of Appendix J requires that test intervals for Type A, Type 8, and Type C testing be determined by using a performance-based approach. Currently, the Surry 10 CFR 50 Appendix J Testing Program Plan is based on RG 1.163, which endorses NEI 94-01, Revision 0. This LAR proposes to revise the Surry 10 CFR 50, Appendix J Testing Program Plan by implementing the guidance in NEI 94-01, Revision 3-A.

By letter dated June 25,2008, (Accession No. ML081140105), the NRC published a SE, with limitations and conditions, for NEI 94-01, Revision 2. In theSE the NRC concluded that NEI 94-01, Revision 2, describes an acceptable approach for implementing the optional performance-based requirements of 10 CFR 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.

At Surry these tests were performed for Unit 1. in 2006, and for Unit 2 in 2000, at a pressure of 44.46 psig, this pressure is higher than the peak calculated design basis loss-of-coolantaccident (LOCA) pressure, Pa, which for Surry is 43.95 psig. In its March 25, 2014, Request for Additional

.Information (RAI) response, the licensee $tated that Table 5.4-11 in the Updated Final Safety Analysis Report (UFSAR) Revision 45, which was transmitted to the NRC by letter dated September 30, 2013, currently reflects a peak pressure value of 44.05 psig. The UFSAR will be revised in the next submittal to the NRC to reflect the recently recalculated value of 43.95 psig.

The leakage rate testing requirements of 10 CFR 50 Appendix J Option B (Type A, Type B and Type C Tests) and the CISI requirements mandated by 10 CFR 50.55a, together, ensure the continued leak-tight and structural integrity of the containment during its service life.

The licensee stated that it meets the limitations and conditions of the SEs for both NEI 94-01,

_Revision 2 and Revision 3-A. NEI 94-01, Revision 2, was issued in 2008, and included provisions for extending the ILRT Type A intervp.l to 15 years subject to the limitations and conditions provided in the SE for Re\(ision 2. Revision 3-A was issued in July 2012, and included guidance for extending the Type C LLRT interval to 75 months. 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 primary containment by minimizing potential leakage paths.

The NRC staff finds that the use of NEI 94-01, Revision 3-A, is acceptable for referencing by.

licensees proposing to amend their TSs to permanently extend the ILRT surveillance interval to 15 years, provided the following applicable conditions are satisfied:

Condition 1 For calculating the Type A leakage rate, the licensee should use the definition in the NEI TR 94-01, Rev. 2, in lieu of that in American National Standards Institute/American Nuclear Society (ANSI/ANS) ANSI/ANS-56.8-2002.

Staff's Assessment The licensee addressed Condition 1 in its August 12, 2013 letter. Surry usesthe definition found in Section 5.0 of NEI 94-01, Revision 3-A, for calculating the Type A leakage rate (GTP-315 Step 4.1.4.A). NEI 94-01, Revision 3-A, contains the same definition as NEI 94-01, Revision 2.

The licensee stated that the ILRT testing history of the containment structure leakage is acceptable, with margin, and no failed ILRTs.

The TS acceptance criterion for maximum allowable containment leakage rate, La, at Pa, is 0.1 percent(%) by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The last two consecutive successful tests were performed in April 1992, and May 2006, for Unit 1, and May 1991, and

October 2000, tor Unit 2. The licensee stated that the value of Pa for Surry is 45.1 psig per TS 6.8.4(g).

The results provided show that the two most recent tests performed in 1992 and 2006 for Unit 1, and 1991, and 2000 for Unit 2, were successful with containment performance leakage rates less than the maximum allowable containment leakage rate (1.0 La at Pa) of 0.1% containment air weight per day, thus the NRC staff finds that, consistent with the guidance in NEI 94-01, Revision 3-A, this performance history for Type A tests supports extending the current ILRT interval to 15 years.

SURRY ILRT RESULTS Completion Type A TypeC Containment Date 95% Upper Confidence Pathways Leakage Rate Level (UCL) (ILRT)

Unit 1 0.376 of La 0.010 of La 0.386 of La April1992 Unit 1 0.267 of La 0.031 of La 0.298 of La May 2006 Unit 2 0.414. of La 0.004 of La 0.418 of La May 1991 Unit 2 0.050 of La 0.010 of La 0.060 of La October 2000 Condition 2 The licensee submitted a schedule of containment inspections to be performed prior to and

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between Type A tests.

Staff's Assessment The licensee addressed Condition 2 in its August 12, 20131etter. The licensee provided an approximate schedule for the containment surface examinations assuming the Type A test frequency is extended to 15 years. In addition, the licensee stated that Surry has established procedures for performing visual examinations of the accessible surfaces of the containment for detection of structural problems. RG 1.163, Regulatory Position C.3 specifies that these examinations should be conducted prior to initiating a Type A test and during two other outages before the next Type A test if the interval for the Type A test has been extended to ten years, in order to allow for early detection of evidence of structural deterioration. These visual examinations have been completed, with no significant defects noted to date. The licensee also provided an approximate schedule for the containment surface examinations assuming the Type A test frequency is extended to 15 years.

The licensee also stated that the American Society of Mechanical Engineers (ASME)Section XI Program requires that the steel containment vessel be examined in accordance with the requirements of the ASME Boiler and Pressure Vessel Code (B&PV),Section XI, Subsection IWE, and associated modifications and limitatiOQS imposed by 10 CFR 50.55a(b)(2).

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The licensee provided a list of containment inspections to be performed prior to and between Type A tests in Section 4.4 of the LAR. The general visual inspections requirements noted in the LAR, meet the criteria noted in NEI 94-01, Revision 3-A.

Condition 3 The licensee addresses the areas of the containment structure potentially subjected to '

degradation.

Condition 4 The licensee addresses any tests and inspections performed following major modifications to the containment structure, as applicable.

Staff's Assessment of Conditions 3 and 4 In the August 12, 2013, .letter the licensee stated that it has determined that Surry has no areas which require augmented examinations of surface areas likely to experience accelerated degradation and aging. In addition, the licensee stated that general visual examinations of accessible interior and exterior surfaces of the containment system for structural problems are conducted in accordance with the Surry IWE/IWL CISI Plans which implement the requirements of the.ASME,Section XI, Subsections IWE and IWL, as required by 10 CFR 50.55a(g).

In the August 12, 20131etter, the licensee provided a summary of LLRT results which demonstrated acceptable performance history in accordance with the Containment Leakage Rate Program.

In the August 12, 2013 letter, the licensee stated that major modifications to the containment structure containment structure were needed for the steam generators and reactor vessel closure head replacement for both Units 1 and 2. On those occasions, ,the design change process addressed the testing requirements of the containment structure modifications.

Based on the evaluation above, the NRC staff finds that the licensee has effectively addressed Conditions 3; the licensee has identified the areas of containment potentially subjected to degradation, as well as any test or inspections performed following major modifications to the containment. In addition, the licensee has implemented an adequate Containment Leakage Rate Testing ILRT and LLRT program, CISI and supplemental inspections to periodically examine,

• monitor, and.manage age-related and environmental degradation of the Surry primary containment.

Condition 5 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.

Condition 5 is not applicable because the licensee is not requesting to extend the ILRT interval for more than 15 years.

Condition 6 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 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, including the

. use of past containment ILRT data.

Condition 6 is only applicable to plants licensed under 10 CFR Part 52. Surry is not licensed pursuant to 10 CFR Part 52.

Based on the evaluation above, the NRC staff finds that the licensee has effectively addressed all applicable Conditions; the licensee has identified the areas of containment potentially subjected to degradation, as well as any test or inspections performed following major modifications to the containment. In addition the licensee has implemented an adequate Containment Leakage Rate Testing ILRT and LLRT program, CISI and supplemental inspections to periodically examine, monitor, and manage age-related and environmental degradation of the Surry primary containment.

The results of the past ILRTs and LLRTs programs demonstrate acceptable performance of the Surry primary containment and demonstrate that the leak-tight integrity of the primary containment is adequately managed. The leak-tight integrity of the Surry primary containment will continue to be periodically monitored and managed by the LLRT and CISI programs if the current ILRT interval is extended from 10 years to 15 years. Thus, the NRC staff finds that there is reasonable assurance that the containment leak-tight integrity will continue to be maintained, without undue risk to public health and safety if the current ILRT interval at Surry is extended to 1'5

• years. Therefore, the NRC staff finds it acceptable to extend the current ILRT Surry interval at Surry from 10 years to 15 years as proposed by the licensee, in accordance with NEI 94-01, Revision 3-A. The next Type A test for Unit 1 may therefore be conducted no later than May 16, 2021, in lieu of the current due no later than May 6, 2016. The next Type A test for Unit 2 due will be conducted no later than October 26, 2015.

3.1.2:2 Performance-Based Type C Test LLRT Frequencies The reactor containment leakage test program requires the licensee to perform ILRT, also termed as a Type A test, and LLRTs termed as Type Band 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*purpose of NEI TR 94-01, Revision 3~A. is to assist licensees in the implementation of Option B to 10 CFR Part 50, Appendix J, and in extending Type C LLRT intervals beyond

. 60 months. Specifically, NEI TR 94-01, Revision 3-A, includes guidance that would permit licensees to extend the Type C LLRT surveillance intervals to 75 months. It delineates a

\

performance-based approach for determining Type A, Type B and Type C containment leakage rate testing frequencies.

In the SE issued by the NRC dated June 8, 2012, the NRC staff concluded that NEI 94-01, Revision 3-A, as modified to include two limitations and conditions, is acceptable for referencing by licensees proposing to amend their TS with regard to containment leakage rate testing for the extension of Type C testing from 60 months to 75 months.

The following addresses the two Conditions of the June 8, 2012, NRC SE of NEI 94-01, Revision 3-A.

Condition 1 The NRC staff is allowing the extended interval for Type C LLRTs to 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 NRC staff is also allowing the non-routine emergent extension out to 84 months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI 94-01, Revision 3-A. At no time shall an extension pe allowed for Type C valves that are restricted categorically (e.g. BWR MSIVs), 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.

Staff's Assessment The licensee states in the August 12, 2013, letter that following the approval of the amendment, Surry will follow the guidance of NEI 94-01, Revision 3-A to assess and monitor margin between the Type B and C leakage rate summation and the regulatory limit. This will include corrective actions to restore margin to an acceptable level.

10 CFR Part 50, Appendix J, states that Type C tests shall be performed prior to initial reactor operation. In accordance with the guidance in NEI 94-01, Revisipn 3-A, subsequent periodic Type C tests shall be performed at a frequency of at least once per 30 months, until adequate performance history is established. Extensions of Type C test intervals are allowed based upon completion of two consecutive periodic as-found tests where the results of e*ach test are within a licensee's allowable administrative limits.

The total allowable "as-left" leakage from Type B and C components is 0.6 La; for Surry this equates to 174 standard cubic feet per hour (scfh). The licensee provided a summary of the maximum and minimum pathway leak rate for the last three refueling outages for both Units.

During the last two refueling outages for each unit, the Type B and C leakage rates are less than the administrative leakage rate limit with sufficient margin, and are considered acceptable to the staff.

The NRC staff has determined that administrative limits for leakage rates have been established, documented and maintained for each Type C component prior to the performance of LLRT in accordance with the guidance provided in ANSI/ANS-56.8-2002, Sections 6.5 and 6.5.1. The administrative limits are specific to individual penetrations or valves, and not the surveillance

acceptance criteria for Type C tests. Acceptance criteria for the combined leakage rate for all penetration subject to Type C testing is defined in accordance with ANSI/ANS-56.8-2002, Sections 6.4 and 6.5.

Condition 2 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 Types B and C total and 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.

The licensee states in the August 12, 2013, letter that following the approval of the amendment, consistent with the guidance of Section 11.3.2 of NEI 94-01, Revision 3-A, Surry will estimate the amount of understatement in the Types B and C total and include determination of the acceptability in a post-outage report.

As documented in NUREG-1493, industry experience has shown that most ILRT failures result from leakage that is detectable by LLRT (Type Band C testing). Specific testing frequencies for" the LLRTs are reviewed prior to every refueling outage. An outage scope document is issued to document the LLRT periodically and to ensure that all pre-maintenance and post-maintenance testing is complete. The post-outage report provides a written record of the extended testing interval changes c;1nd the reasons for the changes based on testing results and maintenance history. Based on the above measures, the LLRT program will provide continuing assurance that the most likely sources of leakage will be identified and repaired.

ANSI/ANS-56.8-2002, Section 6.4.4, also specifies surveillance acceptance criteria for Type B and Type C tests and states that: "The combined [as-found] leakage rate of all Type B and Type C tests shall be less than 0.6 La when evaluated on a minimum pathway leakage rate basis, at all times when containment operability is required." It states, moreover, that: "The combined leakage rate for all penetrations subject to Type B and Type C test shall be less than or equal to 0.6La as determined on an maximum pathway leakage rate basis from the as-left LLRT results."

These combined leakage rate determinations shall be done with the latest leakage rate test data available, and shall be kept as a running summation of the leakage rates.

The containment components' monitoring and maintenance activities will be conducted according to the requirements of 10 CFR 50, Appendix J, 10 CFR 50.55a, and 10 CFR 50.65 (a). The above provisions are considered to be acceptable performance monitoring strategies for assuring that the risk of the proposed change will remain small.

Based on the above, the staff considers the adoption of Type C 75 month interval for Surry, Units 1 and 2, are acceptable. The important performance factors have been identified, and are considered in establishing testing intervals, which include past performance, service conditions, d~sign, safety impact, and cause determination. The licensee has developed bases for new frequencies based upon satisfactory performance of leakage tests that meet the requirements of

.: 11 -

10 CFR Part 50, Appendix J. The considerations used to determine appropriate frequencies include service life, environment, past performance, design, and safety impact.

3.1.3 Conclusion The NRC staff finds that the information provided.by Dominion in its LAR and supplements address the limitations and conditions to NEI 94-01, Revision 2 and 3-A.

Based on the above evaluation, the NRC staff finds that there are no significant increases in risk or reductions in safety resulting from the requested test extension, beyond those already considered in the establishment of the intervals allowed by RG 1.163 and NEI 94-01, Revision 3-A.

Further, the Surry containment has a good recent leakage rate history. Therefore, the NRC staff concludes that the requested TS change, increasing the Type A test interval frequency permanently to 15 years and increasing the Type C test interval to 75 months, is acceptable.

3.2 Probabilistic Risk Assessment 3.2.1 Background Section 9.2.3.1, "General Requirements for ILRT Interval Extensions Beyond Ten Years," of Nuclear Energy Institute (NEI) 94-01, Revision 2-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," states that plant-specific confirmatory analyses are required when extending the Type A ILRT interval beyond ten years.

Section 9.2.3.4, "Plant-Specific Confirmatory Analyses," of NEI 94-01 states 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 \"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 theSE dated June 25, 2008, the NRC staff found the methodology in NEI 94-01, Revision 2-A, and EPRI TR-1 009325, Revision 2, acceptable for referencing by licensees proposing to amend their TS to permanently extend the ILRT interval to 15 years, provided certain conditions are satisfied. These conditions, set forth in Section 4.2 of the SER for EPRI TR-1 009325, Revision 2, stipulate that:

1. The licensee submit documentation indicating that the technical adequacy of their Probabilistic Risk Assessment (PRA) is consistent with the requirements of RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk .

Assessment Results for Risk-Informed Activities," relevant to the ILRT extension application. Additional application specific guidance on the technical adequacy of a PRA used to extend ILRT intervals is provided in the SER for EPRI TR-1 009325, Revision 2.

1 EPRI TR-1009325, Revision 2-A, is also identified as EPRI TR-1018243. This report is publicly available and can be found at www.epri.com by typing "1 018243" in the search field box.

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 2 of the SER for EPRI TR-1 009325, Revision 2 .

. 3. The methodology in EPRI TR-1 009325, Revision 2, is acceptable provided the average leak rate for the pre-existing containment large leak accidenrcase (i.e.,

accident case 3b) used by licensees is assigned a value of 100 times the maximum allowable leakage rate (La) instead of 35 La.

4. A license amendment request is required in instances where containment over-pressure is relied upon for emergency core cooling system (ECCS) performance. In addition, the change in core damage frequency (CDF) should be calculated and reported.

3.2.2 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 assessment was provided in Attachment 4 of the LAR submitted August 12, 2013 (ADAMS Accession No. ML13232A042).

In Section 4.6.1 of Attachment 1 to the LAR, the licensee stated that the plant-specific risk assessment follows the guidance in NEI 94-01, Revision 3-A3 (ADAMS Accession No. ML12221A202); the methodology described in EPRI TR-1009325, Revision 2-A; and the NRC regulatory guidance outlined in RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to. the Licensing Basis." Additionally, the methodology used for Calvert Cliffs Nuclear Power Plant to assess the risk from undetected containment leaks due to corrosion was used to perform a sensitivity analysis.

The licensee addressed each of the four conditions for the use of EPRI TR-1 009325, 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.

3.2.1 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.

• Consistent with the information provided in Regulatory Issue Summary (RIS) 2007-06 (ADAMS Accession No. ML070650428), "Regulatory Guide 1.200 Implementation," the NRC staff will use Revision 2 of RG 1.200 (ADAMS Accession No. ML090410014) to assess technical adequacy of 2 Section 4.2 of the SER for EPRI TR-1009325, Revision 2, indicates that the clarification regarding small increases in risk is provided in Section 3.2.4.5; however, the clarification is actually provided in Section 3.2.4.6.

3 NEI 94-01, Revision 3-A, added guidance for extending Type C Local Leak Rate Test (LLRT) surveillance intervals beyond sixty months. The guidance for extending Type A ILRT surveillance intervals beyond ten years, is the same as that in Revision 2-A.

~ 13-the PRA used to support risk-informed applications received after March 2010. In Section 3.2.4.1 of the Safety Evaluation Report (SER) for NEI 94-01, Revision 2 and EPRI TR-1 009325, Revision 2, the NRC staff states that Capability Category I of the ASME PRA standard shall be applied a.s the standard for assessing PRA quality for IRLT extension applications, since approximate values

• of CDF and large early release frequency (LEAF) and their distribution among release categories are sufficient to support the evaluation of changes to ILRT frequencies.

• Section 4.6.1 of Attachment 1 to the LAR states that:

Ttie current Surry Level 1 and LERF internal events PRA model was used to perform the plant~specific risk assessment. This PRA model has been peer reviewed against the PRA Standard [ASME/ANS RA-Sa-2009] to meet RG 1.200, Revision 2, and gaps between the PRAmodel and PRA standard are addressed as a part of the PRA technical adequacy evaluation discussed in Attachment 5.

Per Attachment 5 to the LAR, the most recent at-power internal events model, referred to as S007 Aa, was used to analyze the risk of the extending the Type A ILRT interval to 15 years for Surry, Units 1 and 2, and the effective date of this model is September 30, 2009. The most recent focused scope peer review was conducted in June 2012, and compared this model against the requirements of the ASME/ANSRA-Sa-2009 PRA standard and the clarifications and qualifications provided in the NRC endorsement of the standard contained in ~G 1.200, Revision 2.

• Attachment 5 also states that:

The PRA model is maintained and updated under a PRA configuration control program in accordance with Dominion procedures. Plant changes, including physical and procedural modifications and changes in performance data, are reviewed and the PRA model is updated to reflectsuch changes periodically by qualified personnel, with independent reviews and approvals.

Section 4.6.2 of Attachment 1 to the LAR reports that "95.5 percent of the [Supporting Requirements] (SRs) were considered met with Category 1/11 or greater" and gaps identified by the peer reviews are evaluated for impact on the ILRT extension application.

Attachment 5 to the LAR provides all open gaps identified in the 1998, 2010, and 2012 peer reviews and evaluates their impact in Tables 1, 3 and 4, respectively. Many facts and observations (F&Os) from the peer reviews have been dispositioned and are considered closed by the licensee. Table 6 in Attachment 5 provides the resolutions to the Category B (i.e.,

important and necessary to address) F&Os from the 1998 Peer Review as well as two F&Os from the 2010 Focused Peer Review which were closed. The PRA model gaps that are considered closed were not evaluated for impact on the application. Many of the open findings identified are considered by the licensee to pertain to documentation issues or enhancements to the PRA model that do not affect CDF or LERF.

In some instances, the licensee discussed the use of sensitivity studies to address potential nonconservatisms in the PRA model which were identified by the remaining open F&Os. In all instances, the. licensee determine.d that the impact on CDF and LERF were minimal and would

not impact the acceptability of the ILRT interval extension risk results. The NRC staff agrees that the estimated risk increases, as reviewed in Section 3.2.2 of this Safety Evaluation, meet the guidelines in RG 1.174 with a reasonable margin such that the cumulative effect of these potential nonconservatisms would have a negligible impact on the ILRT interval extension.

Section 4.6.1 of Attachment 1 to the LAR states that the risk analyses address the dominant

.*external events using conservative expert judgment with the information from the Surry Individual Plant Examination of External Events (IPEEE) and that insights and informationfrom the IPEEE, as updated in 2006, have been used to estimate the effect on total LERF due to these external events. Section 5. 7 of Attachment 4 to the LAR states that "the method chosen to account for external events contributions is similar to the approach used to calculate the change in LERF for the internal events using the guidance" in EPRI TR-1 009325, Revision 2-A. This approach was

. used for external events including internal fires and seismic events. The licensee notes that "external events such as high winds, external floods, transportation, and nearby facility accidents were considered and screened in the IPEEE," so their risk impact was considered to be "negligible compared to the impact associated with internal fires and seismic events."

• Attachment 4 to the LAR also states that:

The IPEEE only evaluated the external events risk associated with Surry, Unit 1. However, it also determined that the differences between Unit 1 and Unit 2 would have negligible impact on the PRA results, so the IPEEE CDF _and LERF was taken as representative of both Unit 1 and Unit 2.

In Section 3.2.4.2 of the SER for NEI 94-01, Revision 2 and EPRI TR-1 009325, Revision 2, the NRC staff states that:

Althoug~ the emphasis of the quantitative evaluation is on the ~isk 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."

Therefore, the information used to estimate the effect on total LERF due to external events is considered acceptable. The risk impact of external events is included in the LAR and the increase .

in LERF was determined to meet the guidelines in* RG 1.174 as discussed in Section 3.2.2 of this

• SE.

Given that the licensee has evaluated its PRA against the current ASME PRA standard and Revision 2 of RG 1.200, evaluated the findings developed during the peer reviews of its PRA for applicability to the ILRT interval extension, and addressed the findings, 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.

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3.2.2 Estimated Risk Increase The second condition stipulates that the licensee submit documentation indicating that the estimated risk increase associated with permanently extending the 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, and EPRI TR-1009325, Revision 2.

Specifically, a small increase in population dose should be defined as an increase in population dose of less than or equal to either 1.0 person-rem per year or 1 percent of the total population dose, whichever is less restrictive. In addition, a small increase in conditional containment failure probability (CCFP) should be defined as a value marginally greater than that accepted in previous one-time 15-year ILRT extension requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage points .. Additionally, for plants that rely on containment over-pressure for net positive suction (NPSH)for ECCS injection, both CDF and LERF will be considered in the ILRT evaluation and compared with the risk acceptance guidelines in RG 1.174.

• 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 4.6.3 of Attachment 1 to the LAR. Details of the risk assessment are provided in Attachment 4. The risk assessment is applicable to both Surry, Unit 1 and Unit 2. The reported risk impacts are based on a change in test frequency from three tests in 10 years (the test frequency under 10 CFR 50 Appendix J, Option A) to one test in 15 years. The following conclusions can be drawn from the licensee's analysis associated with extending the Type A ILRT frequency:

1. The reported increase in CDF is 2.67E-11/year, as based on the internal events model.

Using accident analyses, the licensee determined that only large break LOCA scenarios were subject to containment over-pressure concerns that might affect the CDF. Therefore, no additional events, including external events, need to be evaluated to determine the impact on CDF. This change in internal events risk is considered to be "very small" (i.e.,

below 1E-07/year) per the acceptance guidelines in RG 1.174.

2. The reported increase in LERF is 6.79E-08/year for internal events only. This change in internal events risk is considered to be "very small" (i.e., below 1E-07/year) per the acceptance guidelines in RG 1.174. The reported increase in LERF is 3.25E-07/year for internal and external events combined. The risk contribution from external events includes the effects of internal fires and seismic events, as discussed in Section 3.2.1 of this Safety Evaluation. This change in internal and external events risk is considered to be "small" (i.e., between 1E-06/year and 1E-07/year) per the acceptance guidelines in RG 1.174. Per RG 1.174, an assessment of baseline LERF is required to show that the total '-

LERF is less than 1E-05 per reactor year. Per Section 7.0 of Attachment 4, the total base LERF is estimated to be 3.7E-07/yr. Thus, the new total LERF, given the increase in ILRT interval, would be approximately 6.95E-07/year which is below the total LERF value of 1E-05 per reactor year in RG 1.174.

3. Given a change in Type A ILRT frequency from three in 10 years to once in 15 years, the reported increase in the total population dose is 5.47E-03 person-rem per year, or 0.21 percent of the total population dose. These values are below the values associated with a small increase in populationdose, as provided in EPRI TR-1009325, Revision 2-A,and defined in Section 3.2.4.6 of the NRC SER for NEI 94-01, Revision 2. Thus, this increase

in the total integrated plant risk for the proposed change is considered small and supportive of the proposed change.

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 reports the increase in CCFP for going from a test frequency of three in 10 years to one in 15 years to be 0.93 percent. This is below the guideline value of 1.5 percentage points for a small increase in CCFP, as provided in EPRI TR-1 009325, Revision 2-A, and defined in Section 3.2.4.6 of the NRC SER for NEI 94-01, Revision 2. Thus, the NRC staff finds that the defense-in-depth philosophy is maintained based on the small magnitude of the change in CCFP for the proposed amendment.

Therefore, the NRC staff concludes that the increases in CDF and LERF are small and consistent with the acceptance guidelines of RG 1.174, the increase in the total integrated plant risk for the proposed change is small and supportive of the proposed change, and the defense-in-depth philosophy is maintained based on the small magnitude of the change in the CCFP. Accordingly, the second condition is met.

3.2.3 Leak Rate for the Large Pre-Existing Containment Leak Rate Case The third condition stipulates that in order to make the methodology in EPRI TR-1 009325, Revision 2, acceptable, the average leak rate for the pre-existing containment large leak rate accident case (i.e., accident case 3b) used by the licensees shall be 100 La instead of 35 La. As noted by the licensee in the table in Section 4.6.1 of Attachment 1 to the LAR, the methodology in EPRI TR-1 009325, Revision 2-A, incorporated the use of 100 La as the average leak rate for the pre-existing containment large leakage rate accident case (accident case 3b), and this value has been used in the Surry plant-specific risk assessment. Accordingly, the third condition is met.

3.2.4 Applicability if 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, a LAR is required to be submitted. This SE provides the NRC staff evaluation of the LAR that was submitted and therefore this condition is met. Section 4.6.3 of to the LAR states that Surry Units 1 and 2 credit containment over-pressure to satisfy NPSH requirements for recirculation spray (RS) and low-head safety injection (LHSI) in recirculation mode during LOCAs. The licensee has performed additional accident analyses to address the effects of a containment leak rate of 100 La on available NPSH (discussed in Enclosure A to Attachment 4 of the LAR), and risk analysis to address containment ove-r-pressure impact on CDF, as discussed in Section 5.8 of Attachment 4 of the LAR. According to the clarification provided in Section 3.2.4.6 of the NRC SER for NEI 94-01, Revision 2, and EPRI TR-1 009325, 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 the LAR and the increase in CDF was determined to meet the guidelines in RG 1.174 as discussed in Section 3.2.2 of this SE. Accordingly, the fourth condition is met.

3.2.5 Conclusion Based on the above, the NRC staff concludes that the proposed LAR for a permanent extension of the Type A containment ILRT frequency from once in 10 years to once in 15 years, for Surry Units,.1 and 2, is acceptable. In accordance with revised TS 4.4.8, the next Type A containment ILRT for Surry, Unit 1, shall be performed by no later than May 16, 2021, and the next Type A

. containment ILRT for Surry, Unit 2, shall be performed by no later than October 26, 2015.

3.3 Mechanical & Civil Engineering 3.3.1. Licensee's Proposal for Adoption of NEI 94-01, Revision 3-A In Reference 4.1, the licensee notes that Surry TS 4.4.8, "Containment Leakage Rate Testing Requirements," currently states:

"The containm.ent and containment penetrations leakage rate shall be demonstrated by performing leakage rate testing as required by 10 CFR 50 Appendix J, Option 8, as modified by approved exemptions, and in accordance with the guidelines contained in Regulatory Guide 1. 163, dated September, 1995 as modified by the following exception:

NEI 94-01-1995, Section 9.2.3: The first Unit 2 Type A test performed after the October 26, 2000 Type A test shall be performed no later than October 26, 2015."

The proposed amendment would revise this portion of TS 4.4 by replacing the reference to RG 1.163 with a reference to NEI 94-01, Revision 3-A, as follows:

"The containment and containment penetrations leakage rate shall be demonstrated by performing leakage rate testing as required by 10 CFR 50 Appendix J, Option B, as modified by approved exemptions, and in accordance with the guidelines contained in N£194-01, Revision 3-A, "Industry Guidelines for Implementing Performance-Based Option of 10 CFR 50, Appendix J," dated July 2012."

The NRC final SEdated June 8, 2012 (Reference 4.7), documents the NRC's evaluation and acceptance of NEI 94-01, Revision 3, subject to the specific limitations and conditions listed in Section 4 of the SE. Reference 4.6 describes an approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J. It incorporates the regulatory positions stated in RG 1.163 and includes provisions for extending Type A test intervals up to 15 years and Type C LLRT intervals up to 75 months. In order to use the guidance in Reference 4.6, licensees must address the conditions and limitations in the NRC SE for Revision 2 (Reference 4.5) and Reference 4.7. This issue is discussed further in a letter from the NRC to NEI dated August 20, 2013, which requested NEI to update Revision 3-A to include the conditions in both SEs (Reference 4.8).

The licensee submitted the proposed TS change in accordance with 10 CFR 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."

3.3.1.2 NRC Conditions in NEI 94-01, Revision 2-A (Reference 4.5)

a. Condition 1 - Licensee Proposal 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 ANSI/ANS-56.8-2002.

(Refer to SE Section 3.1.1.1 )."

In Section 4.0 of Reference 4.1, the licensee noted that following NRC approval of this LAR, Dominion will use the. definition in Section 5.0 of NEI 94-01, Revision 3-A, for calculating the Type A leakage rate when future Surry Type A tests are performed. The licensee further noted that the definitions in Revisions 2-A and 3-A are identical. This issue was included as a formal commitment in Attachment 6, "List of Regulatory Commitments" of Reference 4.1.

Condition 1 -NRC Staff Evaluation On the basis that the licensee has committed to comply with the definition in Section 5 of NEI 94-01, Revision 2-A, which is identical to that in Revision 3-A, the NRC staff finds that the licensee has adequately addressed Condition 1 of Reference 4.5.

b. Condition 2 - Licensee Proposal NRC Condition 2 states: "The licensee submits a schedule of containment inspections to be p(:lrformed 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, "Supplemental Inspection Requirements,"

states that in order to provide a "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. It is recommended that these inspections be performed in conjunction, or coordinated, with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Subsection IWE/IWL required examinations."

Section 3.1.1.3 of Reference 4.5 notes that, to avoid duplication or deletion of examinations, licensees using TR NEI 94-01, Revision 2, 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, S~bsection IWE and IWL requirements.

In Section 4.4 of Reference 4.1, the licensee listed the planned visual inspections of exterior surfaces of the containment. The Unit 1 table lists one examination prior to the last Type A test, and then two examinations between the last Type A test and the next scheduled Type A test. The Unit 2 table lists.two examinations between the last Type A test and the next scheduled Type* A test. In Reference 4.2, the licensee noted that the original table only listed visual examinations conducted or scheduled to meet the ASME Code Section XI, Subsection IWE and IWL requirements. Additionally, the licensee added

a column to the original table which identified refueling outages in which general visual examinations of accessible interior and exterior containment surfaces were, or will be performed. The licensee noted that these general visual examinations were conducted in accordance with the requirements of the 10 CFR 50 Appendix J testing program.

Condition 2 ~NRC Staff Evaluation The NRC staff reviewed the updated table provided in the Reference 4.2 and noted that at least three general visual examinations of accessible interior and exterior surfaces of the containment will be conducted between the previous ILRT test and the next scheduled test. The staff also noted that these examinations will be in accordance with the Appendix J testing program. Therefore, on the basis that the licensee's schedule of general visual examinations 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 Reference 4.1 and the supplemental information, meets the general visual

. examination requirements in NEI 94~01, Revision 2-A, and 10 CFR 50 AppendixJ, Option B, and therefore, satisfies Condition 2 of Reference 4.5.

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

Section 3.1.3 of Reference 4.5, in part, notes that licensees referencing NEI 94-01, Revision 2-A, iri support of a request to amend their TS should also explore/consider such inaccessible degradation-susceptible areas in plant-specific inspections, using viable, commercially avai.lable nondestructive examination (NDE) methods (such as boroscopes, guided wave techniques, etc.- see Report ORNUNRC/LTR-02/02, "Inspection of Inaccessible Regions of Nuclear Power Plant Containment Metallic Pressure Boundaries," June 2002 (ADAMS Accession No. ML061230425). The NRC staff's intent in the SE is that licensees should explore and consider NDE techniques, such as those discussed in the reference, as these advanced technologies become commercially available and viable for implementation in the future. While recognizing that these techniques may not be fully commercially viable at the present time, the staff emphasized 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, where the corrosion was initiated at the' inaccessible concrete-steel interface.

In Reference 4.1, the licensee explained that general visual examinations of accessible interior and exterior surfaces of the containment system are typically conducted in accordance with the Containment lnservice Inspection (lSI) Plan, which implements the requirements of ASME Code Section XI, Subsections IWE and IWL. Inaccessible areas are identified and evaluated in accordance with the requirements of 10 CFR 50.55a(b)(2)(ix)(A) and (E). The licensee also noted that during the 2000 refueling

• outages for Units 1 and 2, the containment liner/floor mat interface was inspected and

eValuated. This evaluation included ultrasound testing (UT) thickness measurements of the liner. In addition, several areas at the liner/floor mat interface were excavated in Unit 1 to further assess the condition of the liner below the interface. Based on the results of the UT measurements and the visual exams, the licensee concluded that there was no significant deterioration at the interface. The licensee further noted that inspections performed since that time have not identified any conditions in the accessible area that would suggest liner degradation in the inaccessible areas below the interface. Based on these results, no new technologies have been implemented at Surry to investigate inaccessible areas. However, the licensee actively participates in owners groups, ASME Code committees, and with NEI to maintain cognizance of ongoing developments within the nuclear industry regarding new technologies to inspect containment liners.

In addition to the examinations at the liner/floor mat interface, the license noted that based upon the results of past examinations of the liner and the ASME Code IWE-1241 evaluation, no areas of the containment liner are currently identified for augmented examinations (IWE Examination Category E-C). Any abnormal degradation of the containment structure identified during any time will be entered into the corrective action program for evaluation, consistent with the guidance in Section 9.2.3.3 of Reference 4.6.

Condition 3- NRC Staff Evaluation The NRC staff notes that the licensee examined inaccessible portions of the liner in the past and did not identify significant signs of degradation. The licensee also reviewed past examinations to ensure no areas of the containment liner required augmented examinations under the requirements of ASME Section XI, Subsection IWE. In addition; the licensee actively participates in industry groups to track ongoing technology developments and operating experience that may inform future inspection plans and identify new technologies for investigating inaccessible areas. In addition, any abnormal degradation that was identified with current inspection methods was entered into the corrective action program for evaluation to determine the cause. Therefore, the staff finds that the licensee has adequately addressed the intent of Condition 3 of Reference 4.5.

d. Condition 4- Licensee Proposal 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 Reference 4.1, the licensee noted that the steam generators and Unit 1 and 2 reactor heads were replaced and that the containment structure was modified at that time. During those changes, the repair process addressed the testing requirements of the containment structure modifications. The licensee further noted that no containment modifications that require a Type A test are planned prior to the next required Type A test per the proposed change (fall 2020 for Unit 1 and fall 2015 for Unit 2). Any unplanned modifications to the containment prior to the next scheduled Type A test would be subject to the special testing requirements of Section IV.A of 10 CFR 50, Appendix J.

Condition 4- NRC Staff Evaluation The NRC staff notes that the licensee already completed the common repairs that would require a major containment modification (i.e., steam generator and reactor vessel head replacements) and does not currently have any plans for major modifications. If any major modifications are made to the containment, the licensee noted that the requirements of Section IV.A of 10 CFR 50, Appendix J would be followed for post-repair testing.

Therefore, the staff finds that the licensee adequately addressed the necessary testing required after major modifications and the intent of Condition 4 of Reference 4.5.

e. Condition 5- Licensee Proposal 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)."

In Reference 4.1, the licensee noted that Dominion acknowledges and accepts the NRC staff position in Condition 5, as communicatea to the nuclear industry 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," dated December 8, 2008.

Condition 5- NRC Staff Evaluation The NRC staff notes that the licensee has acknowledged and accepted the NRC staff position discussed in Condition 5 and clarified in RIS 2008-27. The 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 staff will implement the position in RIS 2008-27 in reviewing such license amendment requests. Therefore, the NRC staff finds that the licensee has adequately addressed Condition 5 of Reference 4.5.

f. Condition 6- Licensee Proposal 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-A, and EPRI Report No. 1009325, Revision 2-A, including the use of past containment ILRT

• data." In Reference 4.1, the licensee noted that this condition is not applicable to Surry Units 1 and 2 since the units are not licensed pursuant to 10 CFR Part 52.

Condition 6- NRC Staff Evaluation The NRC staff finds that Surry Units 1 and 2 are currently operating reactors licensed to 10 CFR Part 50, and therefore, Condition 6 of Reference 4.5 does not apply.

3.3.1.3 NRC Conditions in NEI 94-01, Revision 3-A (Reference 4.7)

a. Condition 1 - Licensee Proposal NRC Condition 1 states in part: "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 Band 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), 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."

• In Reference 4.1, the licensee noted that following approval of the amendment, Dominion will follow the guidance in NEI 94-01, Rev. 3-A to assess and monitor margin between the Type Band C leakage rate summation and the regulatory limit. This will include corrective actions to restore margin to'acceptable levels.

Condition 1 -NRC Staff Evaluation The NRC staff notes that Section 12.1 of NEI 94-01, Rev. 3-A requires that the post-outage report include the combined Type B and C leakage summation, the margin between the summation and the regulatory limit, and a summary of planned corrective actions, if an adverse trend is identified in the margin values. Section 10.2 of NEI 94-01, Rev. 3-A clearly notes that the nine month extension is only for non-routine, emergent conditions and does not apply to valves that are restricted or limited to 30 month intervals or the base interval due to unsatisfactory LLRT performance.

The licensee stated it will follow the guidance in NEI 94-01, Rev. 3-A, which contains clear guidance to address Condition 1 of Reference 4.7. Therefore, the NRC staff finds that the licensee has adequately addressed Condition 1.

b. Condition 2 - Licensee Proposal NRC Condition 2 states in part: '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 Band 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 Reference 4.1, the licensee noted that following approval of the LAR, Dominion will follow the guidance in Section 11.3.2 of NEI 94-01, Rev. 3-A to estimate the amount of understatement in :the Type Band C totals and include a determination of the acceptability in a post-outage report.

Condition 2 - NRC Staff Evaluation The NRC staff notes that Section 11.3.2 of NEI 94-01, Rev. 3-A requires the post-outage report include an estimate of the amount of understatement in the minimum pathway Type B and C summation whenever LLRT intervals are routinely scheduled beyond 60 months.* The estimate must include the reasoning and determination of the acceptability of the interval extension, demonstrating that the totals calculated represent the actual leakage potential of the penetrations.

The licensee stated it will follow the guidance in NEI 94-01, Rev. 3-A, which contains clear guidance to address Condition 2 of Reference 4.7. Therefore, the NRC staff finds that the licensee has adequately addressed Condition 2.

3.3.1.4 Conclusion of the Licensee's Adoption of NEI 94-01, Revision 3-A Based on the above evaluation of each condition, the NRC staff determined that the licensee has adequately addressed and satisfied the six conditions in Section 4.1 of the NRC SE for TR NEI 94-01, Revision 2-A, as well as the two conditions in Section 4.0 of the NRC SE for TR NEI 94-01, Revision 3-A (References 4.5 and 4.7). Therefore, the NRC staff finds it acceptable for Surry to adopt TR NEI 94-01, Revision 3-A, as the implementation document in its TS 4.4.B, "Containment Leakage Rate Testing Requirements."

• 3.3.2 Licensee's Proposal for Extension of Type A Test Interval up to 15 Years and Type C Test Interval up to 75 Months for Both Units *

• Per Reference 4.1, the licensee proposed to extend the current performance-based Type A test

• interval to 15 years for both units by adopting Reference 4.6 as the implementation document in TS 4.4.B. This change allows Dominion to conduct the next Unit 1 Type A test by May 2021, in

• lieu of the current due date of May 6, 2016. The licensee also proposed.to extend Type C LLRT inte~vals up to 75 months. The licensee justified the proposed changes by demonstrating adequate performance of the Surry containments based on plant-specific containment leakage testing program results and containment in-service inspection (IWEIIWL) results and supported by a plant-specific risk assessment, consistent with the guidance in Reference 4.6. Reference 4.1, 4.2 and 4.3 were reviewed and evaluated by the NRC staff, as discussed in this section, from the point of deterministic considerations, with regard to containment structural and leak-tight integrity, if the current ILRT interval is extended to 15 years and the Type C LLRT interval is extended to 75 months.

• Description of the Surry Primary Containments The Surry containments are steel-lined, heavily reinforced, soil-supported concrete structure_s on flat base mats. The structures consist of vertical cylindrical walls with hemispherical domes. The inside diameter is 126ft. and the spring line of the dome is 122ft. 1 in. above the top of the foundation mat. The foundation is 10 ft. thick and the base of the foundation is approximately 66 ft. below finished ground grade. The walls are covered with a 3/8 in. steel liner and the doine is covered with a 1/2 in. steel liner.

The containment structures were designed to withstand all loadings and stresses anticipated during the operation of the plant. The steel liner is attached to and supported by the concrete and functions primarily as a gastight membrane. The containments do not require the participation of the liner as a structural component in order to meet the design loads.

The containments were designed for a leakage rate not to exceed 0.1 percent by weight of containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the calculated peak pressure. They were also designed for subatmospheric conditions during normal operation. During power operation, the St,Jrry containments are continuously maintained at a subatmospheric condition. The licensee noted in Reference 4~ 1 that although not as significant as design basis accident pressure, the fact that containment can be maintained subatmospheric provides an additional degree of assurance that the containment is free of significant leak paths.

Surry Type A Test Performance History In Section 4.2 of Reference 4.1 ,the licensee summarized the results of the last three Type A tests for Units 1 and 2. The licensee noted that the results of the last three tests were all less than the 1.0 La acceptance limit (La equals 0.1 percent of containment air weight per day at test pressure).

The licensee further noted that since the last three as-found results were less than 1.0 La, a test frequency of at least once per 15 years would be allowed per Reference 4.6.

Surry Type B Test and Type C Test Performance History In Section 4.3 of Reference 4.1, the licensee described its Type Band Type C leakage rate test program. The program requires testing of electrical penetrations, airlocks, hatches, flanges, and valves, as required by 10 CFR 50, Appendix J, Option B. The combined Type B and Type C leakage limit is 174 standard cubic feet per hour (seth} for both Surry, Unit 1 and Unit 2. The licensee summarized the minimum (as-found) and maximum (as-left) pathway leak rate totals for the last three refueling outages with a Type A test. The results of the last two outages with Type A tests were well below the limit; however, these outages occurred in 2006 and 1992 for Unit 1 and 2000 and 1991 for Unit 2. No recent information was provided on the Type B and Type C test results. In addition, the licensee noted that eight penetrations in Unit 1 and four penetrations in Unit 2 were being tested on an increased frequency due to leakage performance, without any further discussion. In order to assess the proper and effective implementation of the Type Band Type C local leak rate testing program, by email dated December 12, 2013, the NRC staff requested a summary of the combined leakage results from Type B and Type C tests since the last Type A test and a summary table of recent LLRT results of those containment penetrations (including their test schedule intervals} that have not demonstrated acceptable performance

history, in accordance with the containment leakage testing program and a discussion of the causes and corrective actions taken.

In Reference 4.2, the licensee provided tables for Units 1 and 2 which summarized the Type 8 and C as-found and as-left leakage for both Units since the last Type A test. The tables showed that the as-found and as-left leakage rates have been well below the acceptance criteria of 174 seth. Reference 4.2 also identified the number of valves that failed the administrative limit for each test and explained the corrective actions taken to address the failures. The licensee also noted that these valves will require testing during each refueling outage until two successful leakage rate tests are recorded.

• In Reference 4.3, the licensee explained that there were discrepancies in the provided leakage values. In order to resolve the discrepancies, a validation and independent review of the leakage data was performed. During this review additional discrepancies were identified. The licensee further noted that the validated and independently reviewed data provided in Reference 4.3 superseded the values provided previously. To address the cause and occurrence of these discrepancies, the licensee created a corrective action assignment to review the process used to compile 10 CFR 50 Appendix J data and to determine actions to ensure the maintenance of accurate records.

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

In Section 4.4 of Reference 4.1, the licensee described its Containment Inspection Program, both the ASME Code Section XI, Subsection IWE and Subsection IWL portions. In Section 4.4.1, the licensee described the IWE lnservice Inspection Program and noted that containment liner examinations (IWE) will be performed by the required date of April 25, 2014 for Unit 1 and October 19, 2014, for Unit 2, to the requirements of the 2001 Edition through the 2003 Addenda of the ASME Code Section XI. The licensee noted that the IWE general visual examinations will be performed on the accessible surface areas of both containmen! liners. Past inspections of both liners have noted coating degradation associated with mechanical impact damage. For both liners, there have been no indications of coating blisters, and any corrosion identified at damaged coating sites was limited to superficial pitting. Based on an IWE-1241 evaluation* and past inspection results, no areas have been identified, on either containment liner, that required

. augmented examination;\therefore, the remaining examinations will be general visual examinations (IWE Examination Category E-A).

In Section 4.4.2 of Reference 4.1, the licensee described the IWL lnservice Inspection Program and noted that Surry, Units 1 and 2 completed, or are completing, the ASME Code Section XI, Subsection IWL requirements of their second ten-year Containment lnservice Inspection Program. Concrete containment examinations (IWL) were completed for Units 1 and 2 by the required date of August 31, 2011 in accordance with the requirements of the 2001 Edition through the 2003 Addenda of the ASME Code Section XI. These examinations on the concrete exterior were conducted by the Responsible Engineer using visual (VT~3C and VT-1C) methods. The licensee noted that the examinations identified 47 indications on Unit 1; four of which were designated as code repairs, and 44 indications on Unit 2; one of which was designated as a code repair. Five of the indications required excav(ltion and further examination; the remaining indications were deemed to be cosmetic in nature. Finally, the licensee noted that the conditions

identified to date, either alone or combined, do not adversely affect the ability of either of the containment concrete structures to perform their design function.

Reference 4.2 provided additional information on the type of indications identified during the previous examinations. The licensee noted that most indications were minor spalls, pop-outs, or abandoned anchor bolts/anchor bolt holes. The licensee further explained that the guidance in American Concrete Institute (ACI)-349.3, "Evaluation of Existing Nuclear Safety-Related Concrete Structures," was used to establish whether a repair was code required or cosmetic. The licensee listed the five code repairs and provided a brief description for each one along with how the issue was addressed. Most of the repairs were due to the discovery of embedded materials or loose or hollow sounding concrete. All of the code indications were excavated, inspected by qualified inspectors to ensure the adequacy of the remaining concrete, repaired per code requirements, and re-inspected upon completion of the repair. Per Reference 4.3, the licensee clarified that the values in Reference 4.2 were accurate (i.e., three Unit 1 and two Unit 2 Code repairs) and superseded the values originally reported in Reference 4.1.

3.3.2.1 NRC Staff's Overall Evaluation of the Proposed Extension of Type A Test Interval up to 15 Years and Type C Test Interval up to 75 Months for Both Units The NRC staff reviewed the information related to the licensee's proposal to extend 10 CFR 50, Appendix J test intervals, including leakage test results and ASME Code inspection results. The results provided in Section 4.2 of Reference 4.1 indicate that the previous three consecutive Type A tests at Surry, Units 1 and 2, were successful with containment performance leakage rates less than the maximum allowable containment leakage rate of 0.1 percent containment air weight per day (1.0 La at Pa). Therefore, the NRC staff finds that the performance history of Type A tests supports extending the current ILRT interval to 15 years.

The NRC staff reviewed the local leak rate summary tables provided in Reference 4.2 and the updated results provided in Reference 4.3 and noted that the results for all the recent Type B and C tests were well below the acceptance criteria. The NRC staff reviewed the corrective actions taken for the valves that failed the most recent tests, and noted that the valves would be tested until two successful tests were recorded. The NRC staff also noted that the licensee created an action to investigate the cause of the discrepancies noted in Reference 4.2 and to determine actions that will help ensure accurate record maintenance in the future. Therefore, since the discrepancies from References 4.1 and 4.2 were adequately explained, the Type B and C test ,

results were below the acceptance limit, and the licensee appropriately addressed poor performing valves, the NRC staff finds that the licensee is effectively implementing the Type B and Type C leakage rate test program, as required by 10 CFR 50, Appendix J, Option B.

Additionally, the NRC staff reviewed the ASME Code Section XI, Subsection IWE information and

• noted that the examinations have been completed successfully, with no significant indications of degradation. The NRC staff reviewed the ASME Code Section XI, Subsection IWL information provided in References 4.1 and 4.2, and noted that the licensee had reasonable acceptance criteria in place for identifying code repairs, based on ACI-349.3, which is one of the recommended guidance documents in the ASME Code Section XI, Subsection IWL. In addition, the NRC staff noted that the five ASME Code repairs were properly conducted and inspected by qualified inspectors. Furthermore, the NRC staff noted that the discrepancies identified in References 4.1 and 4.2 were fully clarified in Reference 4.3. Since the submittal discrepancies

were clarified, the ASME Code indications were summarized and none were found to be significant, and the licensee explained how concrete indications were identified for ASME Code repair, the NRC staff finds that the licensee is effectively implementing the ASME Code Section XI, Subsection IWL program. Based on acceptable results of recent ASME Code Section XI, Subsection IWE and Subsection IWL inspections, the NRC staff finds that there was no evidence to date of significant degradation and that the licensee is adequately implementing its containment inspection program to monitor and manage degradation of both containment structures.

In summary, the NRC staff finds that the licensee adequately implemented its Containment Leakage Rate Testing Program (Type A, B, and. C leakage tests), its Containment lnservice Inspection Program, arid necessary supplementary inspections to periodically* examine, monitor, and manage age-related and environmental degradation of the containment structures. The results of past ILRTs, recent LLRTs and the containment concrete and liner visual inspections demonstrate acceptable performance of the containments and demonstrate that the structural and leak-tight integrity of the containment structures were adequately maintained. Thus, the NRC staff determined that there is reasonable assurance that the structural and leak-tight integrity, for

• both containments, will continue to be maintained, without undue risk to public health and safety, if the current Type A intervals are extended to 15 years *and the Type C intervals are extended to 75 months.

3.3.2.2 Conclusion Based on the NRC staff review of the licensee's. submittal of August 12, 2013 (Reference 4.1 ),

supplemental information provided in References 4.2 and 4.3, and the regulatory and technical evaluations above, the NRC staff finds that there is reasonable assurance that the licensee has addressed the NRC conditions to demonstrate acceptability of adopting topical report NEI94-01, Revi!:)ion 3-A. The NRC staff also determined that the structural and leak-tight integrity of the Surry containments will continue to be monitored and maintained, if the performance-based Type A test intervals are extended up to 15 years and the performance-based Type C test intervals are extended up to 75 months. Therefore, the NRC staff concludes that it is ac<;:eptable for Surry Units 1 and 2 to: (i) revise TS 4.4.B, "Containment Leakage Rate Testing Requirements," to adopt NEI 94-01, Revision 3-A, as the implementation document, (ii) extend the current Type A test interval up to 15 years, and (iii) extend the current Type G test intervals up to 75 months.

4.0 REFERENCES

4.1 Grecheck, Eugene S., Letter dated August 12, 2013, Virginia Electric and Power Company to USNRC regarding the Change to Extend Integrated Leak Rate Test Frequency to 15 Years, (ADAMS Accession No. ML13232A042).

4.2 Sartain, Mark D., Letter dated January 24, 2014, Virginia Electric and Power Company to USNRC regarding the Response to Request for Additional Information for Proposed License Amendment Request Permanent Fifteen-Year Type A Test Interval, (ADAMS Accession No. ML14035A219);

• 4.3 Sartin, Mark D., Letter dated March 13, 2014, Virginia Electric and Power Company to USNRC regarding Response to Request for Clarification Proposed License Amendment

Request Permanent Fifteen-Year Type A Test Interval, (ADAMS Accession No. ML14079A082).

4.4 Nuclear Energy Institute Topical Report NEI94-01, Revision 2-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR_Part 50, Appendix J," October 2008 (ADAMS Accession No. ML100620847) 4.5 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).

4.6 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). .

4.7 NRC Final Safety Evaluation Report, "Final Safety Evaluation of Nuclear Energy Institute (NEI) Topical Report 94-01, Revision 3, 'Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, US Nuclear Regulatory Commission, Washington, DC, June 8, 2012 (ADAMS Accession No. ML121030286).

4.8 Bahadur, Sher, Letter dated August 20, 2013, USNRC to Biff Bradley, NEI, to 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, (ADAMS Accession No. ML13192A394).

5.0 STATE CONSULTATION

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

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (78 FR 64548, dated October 29, 2013). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

7.0 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors:

Roberto L. Torres, NRRJDSS/SCVB David Gennardo, NRRIDRA/APLA Bryce Lehman, NRRIDE/EMCB Date: July 3, 2014 -

ML14148A235 *SE input by e-mail OFFICE LPL2-1/PM LPL2-1/LA DSS/SCVB DRA/APLA NAME SWilliams SFigueroa RDennig* HHamzehee*

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!OFFICE DE/EMCB OGC INLO LPL2-1/BC LPL2-1/PM

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