ML18337A422

Issuance of Amendment No. 265 One-Time Extension of 10 CFR Part 50, Appendix J, Type a, Integrated Leakage Rate Test Interval

ML18337A422
Person / Time
Site:	Indian Point
Issue date:	12/20/2018
From:	Richard Guzman Plant Licensing Branch 1
To:	Entergy Nuclear Operations
Guzman R, NRR/DORL/LPL1
References
EPID L-2017-LLA-0406
Download: ML18337A422 (25)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 20, 2018 Vice President, Operations Entergy Nuclear Operations, Inc.

Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 - ISSUANCE OF AMENDMENT NO. 265 RE: ONE-TIME EXTENSION OF 10 CFR PART 50, APPENDIX J, TYPE A, INTEGRATED LEAKAGE RATE TEST INTERVAL (EPID L-2017-LLA-0406)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 265 to Amended Facility Operating License No. DPR-64 for the Indian Point Nuclear Generating Unit No. 3 (Indian Point 3). The amendment consists of changes to the Technical Specifications in response to your application dated December 8, 2017, as supplemented by letter dated July 3, 2018.

The amendment revises Technical Specification 5.5.15, "Containment Leakage Rate Testing Program," to extend the frequency of the primary containment integrated leak rate test, or Type A test, at Indian Point 3. Specifically, the amendment allows for a one-time extension of the integrated leak rate test frequency from 15 years to no later than the plant restart after the Indian Point 3 Spring 2021 (3R21) Refueling Outage (i.e., approximately 16 years).

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

Docket No. 50-286

Enclosures:

1. Amendment No. 265 to DPR-64

2. Safety Evaluation cc: Listserv Sincerely, Richard V. Guzman, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR INDIAN POINT 3, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-286 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 AMENDMENT TO AMENDED FACILITY OPERATING LICENSE Amendment No. 265 License No. DPR-64

1.

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

A.

The application for amendment by Entergy Nuclear Operations, Inc. (ENO, the licensee) dated December 8, 2017, as supplemented by letter dated July 3, 2018, 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.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Amended Facility Operating License No. DPR-64 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. 265, are hereby

.incorporated in the License. ENO shall operate the facility in accordance with the Technical Specifications.

3.

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

Attachment:

Changes to the Amended Facility Operating License and Technical Specifications Date of Issuance: December 20, 2018 FOR THE NUCLEAR REGULATORY COMMISSION

~

c:;~~

James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NO. 265 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 AMENDED FACILITY OPERATING LICENSE NO. DPR-64 DOCKET NO. 50-286 Replace the following page of the Amended Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Page 3

Insert Page 3

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

Remove Page 5.0 - 30 Insert Page 5.0 - 30

(4)

(5) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus or components.

ENO 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.

Arndt. 203 11/27/00 Arndt. 203 11/27/00 C.

This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR D.

E.

Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

( 1)

Maximum Power Level ENO is authorized to operate the facility at steady state reactor core power levels not in excess of 3216 megawatts thermal (100% of rated power).

(2)

Technical Specifications (3)

(4)

The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. 265, are hereby incorporated in the License.

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

(DELETED)

Arndt. 205 2-27-01 (DELETED)

Arndt. 205 2-27-01 (DELETED)

Arndt. 46 2-16-83 (DELETED)

Arndt. 37 5-14-81 F.

This amended license is also subject to appropriate conditions by the New York State Department of Environmental Conservation in its letter of May 2, 1975, to Consolidated Edison Company of New York, Inc., granting a Section 401 certification under the Federal Water Pollution Control Act Amendments of 1972.

Amendment No. 265

Programs and Manual 5.5 5.5 Programs and Manuals 5.5.15 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with NEI 94-01, Revision 2A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," October 2008, as modified by the following exceptions:

a.

ANS 56.8-2002, Section 3.3.1: WCCPPS isolation valves are not Type C tested.

b.

The next Type A test to be performed after the March 2005 Type A test shall be performed no later than the plant restart after the Spring 2021 (3R2I) RFO.

The maximum allowable primary containment leakage rate, La, at a minimum test pressure equal to Pa, shall be 0.1 % of primary containment air weight per day. Pa is the peak calculated containment internal pressure related to the design basis accident.

Leakage acceptance criteria are:

a.

Containment leakage rate acceptance criterion is::: 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria ares_ 0.60 La for the Type Band C tests and_::: 0.75 La for Type A tests;

b.

Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is::: 0.05 La when tested at ~ Pa,

2) For each door, leakage rate is::: 0.01 La when pressurized to.:::: Pa,

c.

Isolation Valve Seal Water System leakage rate acceptance criterion is:::

14,700 ~/hr at> 1.1 Pa.

d.

Acceptance criterion for leakage into containment from isolation valves sealed with the service water system is::: 0.36 gpm per fan (continued)

INDIAN POINT 3 5.0 - 30 Amendment 265

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 265 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-64 ENTERGY NUCLEAR INDIAN POINT 3, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

1.0 INTRODUCTION

By letter dated December 8, 2017, as supplemented by letter dated July 3, 2018 (Agencywide Documents Access and Management System (ADAMS) Package Accession Nos. ML17349A131 and ML18193A465), Entergy Nuclear Operations, Inc. (Entergy, the licensee), submitted a license amendment request (LAR) to revise the Technical Specifications (TSs) for Indian Point Nuclear Generating Unit No. 3 (Indian Point 3). The supplement dated July 3, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or the Commission) staff's original proposed no significant hazards consideration determination as published in the Federal Register on March 13, 2018 (83 FR 10916).

The amendment revises TS 5.5.15, "Containment Leakage Rate Testing Program," to extend the frequency of the primary containment integrated leak rate test (ILRT), or Type A test, at Indian Point 3. Specifically, the amendment allows for a one-time extension of the ILRT frequency from 15 years to no later than the plant restart after the Indian Point 3 Spring 2021 (3R21) Refueling Outage (RFO) (i.e., approximately 16 years).

In its December 8, 2017, LAR, the licensee stated that Indian Point 3 operates on a 24-month refueling cycle, and the next ILRT is required to be performed approximately 1 year prior to the 15th year anniversary of the completion of the last ILRT (March 2005). The licensee also indicated that it entered into a settlement agreement with the Attorney General of the State of New York and Riverkeeper, Inc. to shut down Indian Point 3 by April 30, 2021, subject to operating extension through, but not beyond, 2025 (Entergy letter dated February 8, 2017 (ADAMS Accession No. ML17044A005)).

The licensee further stated that the request to defer the Type A ILRT by approximately 12 months is based on the site's performance history, historical plant-specific containment leakage testing program results, containment inservice inspection program results, and a supporting plant-specific risk assessment.

2.0

2.1 REGULATORY EVALUATION

Description of Containment The Indian Point 3 Updated Final Safety Analysis Report, Subsection 5.1.2, "Containment System Structure Design," states, in part:

The reactor containment structure is a reinforced concrete vertical right cylinder with a flat base and hemispherical dome. A welded steel liner with a minimum thickness of 1/4 inch is attached to the inside face of the concrete to insure a high degree of leak-tightness. The design objective of the containment structure was to contain all radioactive material which might be released from the core following a Loss-of-Coolant Accident. The structure serves as both a biological shield and a pressure container.

The structure,[... ] consists of side walls measuring 148 feet from the liner on the base to the springline of the dome, and has an inside diameter of 135 feet. The side walls of the cylinder and the dome are 4'-6" and 3'-6" thick, respectively.

The inside radius of the dome is equal to the inside radius of the cylinder so that the discontinuity at the springline due to the change in thickness is on the outer surface. The flat concrete base mat is 9 feet thick with the bottom liner plate located on top of this mat. The bottom liner plate is covered with 3 feet of concrete, the top of which forms the floor of the Containment. [... ].

The basic structural elements that were considered in the design of the containment structure are the base slab, side walls and dome acting as one structure under all possible loading conditions. The liner is anchored to the concrete shell by means of stud anchors. The reinforcing in the structure exhibits a total elastic response to all primary loads. The lower portion of the cylindrical liner is insulated to avoid thermal deformation of the liner under accident conditions.

The containment structure is inherently safe with regard to common hazards such as fire, flood and electrical storm. Internal structures consist of equipment supports shielding, reactor cavity and canal for fuel transfer, and miscellaneous concrete and steel for floors and stairs. All internal structures are supported on the 2'-8" thick floor slab.

A 3-foot thick concrete ring wall serving as a missile and partial radiation shield surrounds the Reactor Coolant System components and supports the polar-type reactor containment crane. A 2-foot thick reinforced concrete floor covers the Reactor Coolant System with removable gratings in the floor provided for crane access to the Reactor Coolant Pumps. The four steam generators, pressurizer and various pipings penetrate the floor. Spiral and scissor stairs provide access to the areas below the floor. There is a reinforced concrete missile shield wall around the pressurizer above the operating floor. The original design is to protect the containment steel liner from postulated valve piece or instrument missiles connected to the pressurizer. Currently these missiles have been shown not to be credible.

The refueling canal connects the reactor cavity with the fuel transport tube to the spent fuel pool. The floor and walls of the canal are concrete, with wall and shielding water providing the equivalent of 6 feet of concrete. The floor is 4-feet thick. The concrete walls and floor are lined with 1/4-inch thick stainless steel plate. The linings provide a leakproof membrane that is resistant to abrasion and damage during fuel handling operation.

A sub-surface drainage system is provided around the Containment Building where the mat is below grade (... ). Since the containment is above the water table, no hydrostatic seepage will occur.

2.2 Applicable Regulation and Guidance The regulations in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.54(0) require that the primary reactor containments for water cooled power reactors shall be subject to the requirements set forth in 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." Appendix J to 10 CFR Part 50 includes two options: "Option A - Prescriptive Requirements," and "Option B - Performance-Based Requirements," either of which can be chosen for meeting the requirements of Appendix J.

The testing requirements in 10 CFR Part 50, Appendix J, ensure that: (a) leakage through containments or systems and components penetrating containments does not exceed allowable leakage rates specified in the TSs, and (b) integrity of the containment structure is maintained during the service life of the containment.

Option B of 10 CFR Part 50, Appendix J, specifies performance-based requirements and criteria for preoperational and subsequent leakage rate testing. These requirements are met by performing a Type A test to measure the containment system overall integrated leakage rate of the primary containments; Type B, consisting of a pneumatic test to detect and measure local leakage rates across pressure retaining leakage limiting boundaries; and Type C, consisting of a pneumatic test to measure containment isolation valve leakage rates. After the preoperational tests, these tests are required to be conducted at periodic intervals based on the historical performance of the overall containment system (for Type A tests), and based on the safety significance and historical performance of each penetration boundary and isolation valve (for Type Band C tests) to ensure integrity of the overall containment system as a barrier to fission product release.

The overall integrity (structural and leaktight integrity) of the primary containment is verified by a Type A ILRT, and the integrity of the penetrations and isolation valves is verified by Type Band Type C local leak rate tests (LLRT), as required by 10 CFR Part 50, Appendix J. These tests are performed to verify the essential leaktight characteristics of the containment structure at the design-basis accident pressure. The Type A test also provides a verification of structural integrity.

The leakage rate test results must not exceed the allowable leakage rate (La) with margin, as specified in the TSs. Option B also requires that a general visual inspection of the accessible interior and exterior surfaces of the containment system for structural deterioration that may affect the containment leaktight integrity must be conducted prior to each Type A test, and at a periodic interval between tests.

Section V.B.3 of 10 CFR Part 50, Appendix J, Option B, requires that the regulatory guide (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. Further, the submittal for TS revisions must contain justification, including supporting analyses, if the licensee chooses to deviate from methods approved by the NRC and endorsed in an RG.

The NRC staffs final safety evaluation (SE) for Nuclear Energy Institute (NEI) Topical Report (TR) 94-01, Revision 2, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," and Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2, dated August 2007, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," dated June 25, 2008 (ADAMS Accession No. ML081140105), was incorporated into NEI 94-01, Revision 2-A, November 19, 2008. NEI 94-01, Revision 2-A, describes an NRG-approved approach for implementing the optional performance-based requirements of Option B described in 10 CFR Part 50, Appendix J, which includes provisions for extending Type A ILRT intervals to up to 15 years, and incorporates the regulatory positions stated in RG 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995 (ADAMS Accession No. ML003740058). NEI 94-01, Revision 2-A, delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. This method uses industry performance, plant-specific data, and risk insights in determining the appropriate testing frequency, and also discusses the performance factors that licensees must consider in determining test intervals.

The NRC staff's final SE dated June 8, 2012 (ADAMS Accession No. ML121030286), of NEI 94-01, Revision 3, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," was incorporated into NEI 94-01, Revision 3-A, dated July 2012.

NEI 94-01, Revision 3-A, documents the NRC staff's evaluation and acceptance of NEI 94-01, Revision 3.

The regulations in 10 CFR 50.55a contain the containment inservice inspection program requirements that, in conjunction with the requirements of Appendix J, ensure the continued leaktight and structural integrity of the containment during its service life.

The regulations in 10 CFR 50.65 state, 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 industrywide operating experience.

The regulations in 1 O CFR 50.36 state that the TSs include items in five specific categories.

These categories include: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operations; (3) surveillance requirements; (4) design features; and (5) administrative controls.

NEI 94-01, Revision 0, specifies an initial test interval of 48 months, but allows an extended interval of 10 years based upon two consecutive successful tests. There is also a provision for extending the test interval an additional 15 months in certain circumstances. A one-time extension to the 10-year ILRT frequency for Indian Point 3 was approved by Amendment No. 206 dated April 17, 2001 (ADAMS Accession No. ML011070447). As a result, the last Type A test was conducted in March 2005, which was 15 years since the previous Type A ILRT completed in December 1990. Both of the two most recent Type A tests at Indian Point 3 were successful; therefore, the current interval requirement would have been 1 O years and required to be performed no later than March 2015.

3.0 TECHNICAL EVALUATION

3.1 Background

By License Amendment No. 256 dated March 13, 2015 (ADAMS Accession No. ML15028A308),

the NRC authorized the licensee to implement 10 CFR Part 50, Appendix J, Option B, for Indian Point 3 Types A, B, and C tests. The amendment allowed the next ILRT for Indian Point 3 to be performed within 15 years from the last ILRT (March 2005). Therefore, the next Type A test was to be performed on or before March 2020. As discussed in Section 1.0 of this SE, the licensee's proposed change would defer the period until the Spring 2021 RFO (3R21 ), at which time Entergy will cease Indian Point 3 power operation in accordance with Entergy letter dated February 8, 2017 (ADAMS Accession No. ML17044A005).

Additionally, License Amendment No. 256 allowed for an ILRT maximum interval of 15 years with provision for a grace period of up to 9 months beyond 15 years under the circumstance that an unforeseen emergent condition exists. NRC SE Section 3.1.1.2, "Deferral of Tests Beyond the 15-year interval," for NEI 94-01, Revision 2, states, in part:

As noted above, Section 9.2.3, NEI TR 94-01, Revision 2, states, "Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per 15 years based on acceptable performance history." However, Section 9.1 states that the "required surveillance intervals for recommended Type A testing given in this section may be extended by up to 9 months to accommodate unforeseen emergent conditions but should not be used for routine scheduling and planning purposes." The NRC staff believes that extensions of the performance-based Type A test interval beyond the required 15 years should be infrequent and used only for compelling reasons. Therefore, if a licensee wants to use the provisions of Section 9.1 in TR NEI 94-01, Revision 2, the licensee will have to demonstrate to the NRC staff that an unforeseen emergent condition exists.

3.2 Licensee's Proposed Changes TS 5.5.15 currently states, in part:

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with NEI 94-01, Revision 2A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," October 2008, as modified by the following exception:

ANS 56.8-2002, Section 3.3.1: WCCPPS isolation valves are not Type C tested.

The proposed changes to TS 5.5.15 will add a new exception to allow for the performance of the next Type A test no later than the Spring 2021 (3R21) RFO, as follows (changes identified in bold):

3.3 A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with NEI 94-01, Revision 2A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," October 2008, as modified by the following exceptions:

a.

ANS 56.8-2002, Section 3.3.1: WCCPPS isolation valves are not Type C tested.

b.

The next Type A test to be performed after the March 2005 Type A test shall be performed no later than the plant restart after the Spring 2021 (3R21) RFO.

Historical Leakage Rate Test Results The licensee provided Indian Point 3 historical results of ILRT Type A tests and Type B and Type C local leak rate tests. In addition, the LAR also included a summary of the Indian Point 3 IWE examination results of the containment metal liner completed during RFO 3R17 (Spring 2013) and RFO 3R 18 (Spring 2015), and the IWL results for the containment concrete visual inspections completed in 2015.

3.3.1 Integrated Leakage Rate (Type A) Test History 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 at Indian Point 3 were performed in December 1990 and March 2005. The licensee stated that the value of Pa for Indian Point 3 is 42.38 pounds per square inch gauge (psig) per TS 5.5.15. In Table 1 below, the licensee provided a summary of Type A ILRT results, which demonstrated that the last two Type A tests had containment performance leakage rates less than the La (1.0 La at Pa) of 0.1 percent containment air weight per day.

Table 1 I d" P. t T A ILRT H" t n 1an om ype IS Ory Date As-Found Leakage Test Pressure (pounds per

(% Containment weight per dav)<3>

square inch absolute (psia))<2>

March 2005 0.0565 60.61 December 1990 0.032 59.49 July 27, 1987 0.34(1) 59.89 Auoust4, 1982 0.034 60.00 Auoust 2, 1978 0.14(1) 60.00 Notes:

(1) The Type A as-found test leakage exceeded La in both the 1978 and 1987 ILRTs. These failures were attributed to individual penetration/containment isolation valve leakage and not the primary containment structure (e.g., "welded steel liner").

(2) The calculated peak containment internal pressure for the design-basis loss-of-coolant accident, Pa, is 42.38 pounds psig. The containment design pressure is 47 psig.

(3) The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.1 percent of primary containment air weight per day.

The last two tests for. the Indian Point 3 primary containment have shown a leakage rate much less (i.e., > 94 percent margin) than the acceptance criterion, La, of 0.1 percent of primary containment air weight per day at a test pressure in excess of Pa. This allows continuation of the 15-year interval in accordance with the guidance of NEI 94-01, Revision 2-A. For the Indian Point 3 primary containment, the margin of the test results relative to the acceptance criterion would support a conclusion that exceeding the performance criterion of 0.1 percent would be very unlikely with implementation of the proposed one-time interval extension.

It should also be noted that for the July 27, 1987, ILRT, the licensee stated there was a leak through the reactor coolant pump seal water return valve, the valve was isolated, and the leakage returned to normal. The higher as-found leakage for the August 2, 1978, ILRT was due to a leak in the #33 and #34 containment fan cooler service water supply and return lines inside containment. The March 2005 and December 1990 ILRT Type A results confirmed that the containment is acceptable with respect to the TS design criterion of 0.1 percent leakage of containment air weight per day at the design-basis loss-of-coolant accident pressure (La). Since these last two Type A as-found test results were less than 1.0 La, a test frequency of at least once per 15 years, in accordance with NEI 94-01, Revision 2-A, is acceptable.

The NRC staff reviewed the information related to the licensee's proposal to extend 1 O CFR Part 50, Appendix J, ILRT Type A test intervals, including historical leakage test results.

In Section 4, "Technical Analysis," of its LAR, the licensee provided test results for the two most recent Indian Point 3 ILRT Type A tests of 1990 and 2005. These test results indicate containment performance leakage rates are much less than the maximum allowable containment leakage rate (La at Pa) of 0.1 percent primary containment air weight per day.

Therefore, the staff finds that the performance history of the Type A tests supports extending the current Indian Point 3 ILRT interval to 16 years.

3.3.2 Local Leak Rate Test (Type Band Type C} History In LAR Table 4-1 the licensee presented the Indian Point 3 Types Band C test combined as-found minimum pathway leakage totals since the last ILRT of Spring 2005. These results are recaptured below:

Date

~s-Found 0.6La (cc/min)

Percent ((As-Percent ((As-Leakage (cc/min)

Found/La) x 100)

Found/0.6La) x 100)

April 2005

~1585 119689 0.208 0.347 Aoril 2007 30352 119689 0.152 0.254 April 2009

~4621 119689 0.224 0.373 Aoril 2011 51878 119689 0.260 0.433 March 2013 36669 119689 0.184 0.306 Aoril 2015 56557 119689 0.283 0.473 Uune 2017 59327 119689 0.297 P.496 The TS 5.5.15 acceptance criterion for combined Types B and C test total is s 0.60 La. As displayed in the above table of test results, there has been at least a 50 percent margin for all "As-Found /0.6La" values since the last ILRT during 2005. These results are supportive of a conclusion that the proposed one-time extension of the ILRT interval for Indian Point 3 would be very unlikely to result in the acceptance criterion of 0.60 La being exceeded.

3.3.3 Containment lnservice Inspection Program The leakage rate testing requirements of 10 CFR Part 50, Appendix J, Option B (Type A ILRT),

and the containment inservice inspection requirements mandated by 1 O CFR 50.55a, together, help ensure the continued leaktight and structural integrity of the containment during its service life. As required by TS 5.5.15, Indian Point 3 is subject to the requirements set forth in 10 CFR Part 50, Appendix J, Option B, which requires that test intervals for Type A ILRT be determined by using a performance-based approach. The Indian Point 3 ILRT program is based on implementation of the guidance in NEI 94-01, Revision 2-A.

In its LAR dated December 8, 2017, the licensee provided a summary of the IWE examination results of the containment metal liner completed during RFOs 3R17 (Spring 2013) and 3R18 (Spring 2015), and the IWL results for the containment concrete visual inspections completed in 2015. The licensee stated that Indian Point 3 is committed to the 2001 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, with the 2002 and 2003 Addenda. In accordance with 10 CFR 50.55a(b)(2)(ix)(E), a general visual examination of the Indian Point 3 containment surfaces must be performed during each Section XI inservice inspection period of the 10-year interval. Examinations completed during the referenced RFOs reported minor visual observations with no areas of suspected damage or deterioration that would impact the structural integrity or leaktightness of the containment liner.

The observations were documented in Condition Report (CR) CR-IP3-2013-01382, and included as part of the licensee's LAR dated February 4, 2014 (ADAMS Accession No. ML14050A383), to extend the containment Type A leak rate testing frequency to 15 years.

Also, as part of an amendment to the license renewal application dated February 8, 2017, the licensee performed a one-time inspection of the containment liner at the junction of the concrete slab at the 46-foot elevation. The observations revealed discoloration and exfoliation of the liner from corrosion that resulted from service water leakage from the containment fan cooling unit piping and aggravated by the process of corrosion under the insulation, although no moisture was noted. The licensee concluded that the condition of the liner was acceptable since the level of corrosion was minimal and did not compromise the integrity of the liner. The results of the 2015 inspection were documented in CR-IP3-2015-01888.

In accordance with the ASME Boiler and Pressure Vessel Code, Section IWL, 2001 Edition, 2003 Addenda, as modified by 10 CFR 50.55a, exterior inspection of the containment building was performed in 2015. The IWL examinations are general visual inspections of Class Capability Category components and the reinforced concrete shell of pressure retaining components, and are performed to identify signs of structural degradation that may affect the structural integrity or leaktightness of the containment, and to identify the required repairs and/or replacement activities to minimize degradation due to environmental conditions and aging. The last two inspections were performed in 2009 and 2015. Since several anomalies were identified during the 2009 inspection, those areas were reexamined again during the 2015 inspection to determine if further degradation had occurred. The 2015 inspection identified typical concrete conditions such as minor cracks and pattern cracking, bugholes, leaching, scaling, and spalling.

Based on the licensee's evaluation, the inspection findings were deemed acceptable by the licensee, thereby not impacting the structural integrity of the containment. The results of the 2015 inspection were documented in Entergy Engineering Report IP-RPT-15-00063, "IP3 ASME XI, IWL Concrete Containment Inspection for 2015," which was provided to the NRC by letter dated December 2, 2016 (ADAMS Accession No. ML16350A005), associated with the license renewal application. The licensee stated that the containment liner areas that had experienced some degradation were identified, analyzed, and repaired, as necessary, to ensure an acceptable containment barrier exists.

3.4 Risk Evaluation In the LAR, the licensee performed plant-specific risk analysis for the one-time extension to 16 years using the same methodology as used for the permanent extension to 15 years.

NEI 94-01 provides methodology for plant-specific risk assessment for permanent extension of the ILRT to 15 years. Section 9.2.3.1, "General Requirements for ILRT Interval Extensions beyond Ten Years," of NEI 94-01, Revision 2-A and Revision 3-A states that plant-specific confirmatory analyses are required when extending the Type A ILRT interval beyond 10 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 EPRI TR-1009325, Revision 2-A, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals" (ADAMS Accession No. ML14024A045), also known as EPRI TR-1018243. 1 In the SE dated June 25, 2008, the NRC staff found the methodology in NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2, acceptable for referencing by licensees proposing to amend their TSs to permanently extend the ILRT interval to 15 years, provided that certain conditions are satisfied. These conditions, set forth in Section 4.2 of the SE for EPRI TR-1009325, Revision 2, stipulate that:

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

1. The licensee submits documentation indicating that the technical adequacy of its 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" (ADAMS Accession No. ML090410014), relevant to the ILRT extension application.

2. The licensee submits documentation indicating that the estimated risk increase associated with permanently extending the ILRT surveillance interval to 15 years is small, consistent with the clarification provided in Section 3.2.4.62 of the SE for EPRI TR-1009325, Revision 2.

3. The methodology in EPRI TR-1009325, Revision 2, is acceptable, provided the average leak rate for the preexisting containment large leak accident case (i.e., accident Case 3b) used by the licensee is assigned a value of 100 times the maximum allowable leakage rate (La) instead of 35 La.

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

The licensee performed a risk impact assessment for extending the Type A or containment ILRT frequency. For the risk assessment, the risk was evaluated considering the changes from the base case of performing three tests in 10 years to the proposed case of performing one test in 16 years. The risk assessment was provided in LAR Enclosure 1 entitled, "Risk Impact of One-Time Extending the ILRT Interval Associated with the Proposed Technical Specification Changes." Additional risk information was provided by the licensee in its letter dated July 3, 2018, in response to NRC requests for additional information (RAls) dated May 23, 2018 (ADAMS Accession No. ML181438679).

In Section 4.0 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; EPRI TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals";

EPRI TR-1018243, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals: Revision 2-A of 1009325";

RG 1.17 4, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

January 2018 (ADAMS Accession No. ML17317A256); and Methodology used for the Calvert Cliffs Nuclear Power Plant to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval (letter dated March 27, 2002 (ADAMS Accession No. ML020920100)).

2 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 (ADAMS Accession No. ML12221A202), added guidance for extending Type C LLRT surveillance intervals beyond 60 months. The guidance for extending Type A ILRT surveillance intervals beyond 1 O years is the same as that in Revision 2-A.

The licensee addressed each of the four conditions described in NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2. A summary of how each condition has been determined to be met is provided in the sections below.

3.4.1 Technical Adequacy of the PRA The first condition in Section 4.2 of the safety evaluation report (SER) for EPRI TR-1009325, Revision 2, stipulates that the licensee submit documentation indicating that the technical adequacy of its PRA is consistent with the requirements of RG 1.200, Revision 2, relevant to the ILRT extension application.

Internal Events PRA Consistent with the information provided in Regulatory Issue Summary 2007-06, "Regulatory Guide 1.200 Implementation," dated March 22, 2007 (ADAMS Accession No. ML070650428),

the NRC staff will use Revision 2 of RG 1.200 to assess the technical adequacy of the PRA used to support risk-informed applications received after March 2010. The ASME/ANS PRA standard is the industry consensus standard for PRAs for internal events, internal flooding, fires, and other external events (i.e., seismic, external flooding, high winds, etc.). In Section 3.2.4.1 of the SE for NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2, the NRC staff states that Capability Category I of the ASME PRA standard shall be applied as the standard for assessing PRA quality for ILRT extension applications, since approximate values of core damage frequency (CDF) and large early release frequency (LERF), and their distribution among release categories, are sufficient to support the evaluation of changes to ILRT frequencies.

As discussed in Section 5.0 to Attachment 1 of the LAR, the Indian Point 3 risk assessment performed to support the ILRT application is based on the current Level 1 and Level 2 PRA model of record that was released in November 2012. The NRC staff previously reviewed the technical adequacy of this Indian Point 3 internal events and internal flooding PRA for permanent extension of the ILRT to 15 years (License Amendment No. 256 dated March 13, 2015) and concluded that the PRA is of sufficient technical adequacy to support the evaluation of changes to the ILRT frequency. Therefore, the NRC staff concludes that the Indian Point 3 internal events and internal flooding PRA is of sufficient technical adequacy to support the one-time ILRT frequency extension to 16 years.

External Events In Section 3.2.4.2 of the SE for NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2, the NRC staff states that:

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

Therefore, the staff's review of the contribution of external events for this application is framed by the context that an order of magnitude estimate for the corresponding risk contribution is sufficient. In Section 5. 7 of Enclosure 1 to the LAR, the licensee performed an assessment of external event contribution. The licensee's risk analysis reflected the contribution from internal fire and seismic events. As further discussed below, the licensee assessed that high winds, external floods, and "other" external events were considered negligible in estimation of the external events impact on the ILRT extension application.

To estimate the fire risk, the licensee used a fire CDF of 2.55E-05/year estimated in the individual plant examination of external events (IPEEE) using the EPRI fire-induced vulnerability evaluation method. To estimate the seismic risk, the licensee used the seismic probabilistic risk assessment (PRA) developed during the IPEEE. The licensee stated that after the IPEEE, the seismic PRA CDF has been reevaluated to reflect updated random component failure probabilities and to model recovery of onsite power and local operation of the turbine-driven auxiliary feedwater pump. The licensee used the updated seismic CDF of 2.65E-05/year for the ILRT application. The licensee further stated that a more recent evaluation of seismic risk (Entergy letter dated June 26, 2013 (ADAMS Accession No. ML13183A279)) indicates that the seismic CDF is less than 1 E-5/year. The NRC staff concludes that the licensee's estimate of fire and seismic risk represents an order-of-magnitude estimate, and is, therefore, acceptable for the ILRT application.

As stated in Section 5. 7 of Enclosure 1 to the LAR, the licensee evaluated high winds, floods, and "other" external events based on the Indian Point 3 IPEEE analysis performed in 1997. The licensee concluded that these events are considered negligible in estimation of the external events impact on the ILRT extension risk assessment. In RAl-01, the NRC staff requested justification for the applicability of the IPEEE conclusions to the current plant and its environs, considering each of the external hazards screened from this application and taking into account any updated risk studies and insights. In response to RAl-01, the licensee assessed the applicability of these conclusions to the current as-built and as-operated plant, and any changes in the plant or its environs since the IPEEE analysis was performed. The licensee assessed high winds and tornadoes, external flooding, chemical hazards, transportation and nearby facility accidents, aircraft hazards, and ice. The licensee concluded that the risk from these hazards remains negligible. Because the licensee assessed the effects of other external hazards using the latest available information and concluded minimal quantitative impact, the NRC staff finds this acceptable for the application.

Based on the above considerations, the NRC staff concludes that the internal events, including internal flooding, PRA model used by the licensee is of sufficient technical adequacy to support the evaluation of changes to ILRT frequencies. In addition, the assessments of external events risk are sufficient for application to the ILRT extension evaluation, as they support an order of magnitude estimate consistent with EPRI guidance. Accordingly, the first condition of the NRC staff's SE for EPRI TR-1009325, Revision 2, is met.

3.4.2 Risk Analysis In Section 5.4 of Attachment 1 to the LAR the licensee assessed the internal events contribution to risk, consistent with the methodology in EPRI TR-1009325, Revision 2, as accepted by the NRC in its SE for NEI 94-01, Revision 2, dated June 25, 2008.

In Section 5. 7 of Attachment 1 to the LAR, the licensee performed an assessment of external event contribution. The licensee used two approaches: (1) the multiplier approach, which simply applies a multiplier to the internal events results (the multiplier is derived from the ratio of external events CDF to the internal events CDF), and 2) a second approach where each EPRI accident class frequency is individually estimated. The licensee calculated a total (internal and external events) increase in population dose risk from changing the ILRT frequency from three in 10 years to once in 16 years as 3.67 person-rem/year, or 1.00 percent (when using the multiplier approach), and 3.84 person-rem/year or 0.69 percent (when using the second approach). Because the increase in total population dose reported in the LAR is close to the acceptance criteria values of 1 person-rem/year, or 1 percent, in Section 3.2.4.6 of the NRC SE for NEI 94-01, Revision 2, the NRC staff requested in RAl-02 further justification of the licensee's assumptions. The NRC staff noted that the reduction in the percent population dose change obtained through the use of the second approach relied on increasing the total estimated population dose due to external hazards for the base case, corresponding to the three in 10-year frequency. The NRC staff found that two assumptions appeared to be key to this reduction in population dose change:

• The frequency for the EPRI accident Class 7 sequences (accidents involving containment failure induced by severe accident phenomena) is increased by assuming that 50% of the CDF is due to late Class 7 sequences, as compared to the internal events value of 15%, with the justification that "the external events contributors are dominated by unrecoverable SBO [station black-out]-like scenarios."

• The frequency of the EPRI accident Class 2 sequences (containment isolation failures) is increased by assuming that 0.1 % of the external events CDF is due to large containment isolation failures, as opposed to the internal events contribution of 0.03%, with the justification that "seismic and fire-initiated events would likely be more susceptible" to large containment isolation failures.

In response to PRA-RAl-02, the licensee stated that the key assumption used in the second approach is that the risk profile of the external events scenarios is skewed more toward late containment failures and containment isolation failures than the internal events accident scenarios and addressed the key assumptions as discussed below.

For the first uncertainty item related to the frequency of the EPRI accident Class 7 sequences, the licensee assessed in response to RAl-02 that for seismic events, the frequency assigned to late sequences represented 61.6 percent of the seismic CDF, based on the IPEEE seismic containment event tree. For the fire events, the licensee stated that the IPEEE fire analysis did not include a containment event tree quantification. Therefore, the licensee performed a review of the largest contributors of the fire CDF and, based on this review, provided two cases to estimate the percent of the fire CDF that is binned to the "late" containment failure mode: "Fire Case A" where a human error probability of 0.9 is used for the manual alignment of 480 volt power to support the fan cooler units, and "Fire Case B" where the human error probability is assumed to be 0.0 (the operator never fails to recover containment heat removal). The licensee stated that these cases approximate an upper and lower bound of the percentage of fire scenarios which can lead to late containment failure.

For the second uncertainty item related to the frequency of EPRI accident Class 2 sequences, the licensee stated in response to RAl-02 that, for fire and seismic initiators, large isolation failures can be caused by the failure of the support systems required to close the containment isolation valves, and these failures are not represented in the internal events model. The licensee stated that because the IPEEE did not provide quantitative information to allow a change to the fraction of CDF that would lead to containment isolation failures for seismic and fire events, the updated analysis retained the internal events value of 0.03 percent of the CDF to account for containment isolation failures for seismic and fire events. The NRC staff finds the licensee's assumption of 0.03 percent acceptable for the application.

With the updated assumptions in response to RAl-02, the licensee estimated an increase in the total population dose for a change in the Type A ILRT frequency from three in 10 years to once in 16 years to 3.9 person-rem/year or 0.57 percent (Case A) and 3.81 person-rem/year or 0.80 percent (Case B).

Because the licensee assessed the uncertainties related to the estimate of the population dose change and showed that the percent increase of total population dose remains below the values associated with a small increase in population dose, as provided in EPRI TR-1009325, Revision 2-A, and defined in Section 3.2.4.6 of the NRC SE for NEI 94-01, Revision 2, the NRC staff finds the licensee's updated estimate of external events acceptable for the application.

3.4.3 Estimated Risk Increase The second condition of the NRC staff's SE for EPRI TR-1009325, Revision 2, 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 3.2.4.5 of the NRC SE 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 percent. Additionally, for plants that rely on containment over-pressure for net positive suction head for ECCS injection, both CDF and LERF will be considered in the ILRT evaluation and compared with the risk acceptance guidelines in RG 1.17 4. As discussed in Section 3.2.4 of this SE, Indian Point 3 does not credit containment over-pressure for ECCS performance; therefore, CDF is not a relevant risk metric for this application. Thus, the associated risk metrics include LERF, population dose, and CCFP.

The licensee reported the results of the plant-specific risk assessment in Section 4.0 of to the LAR. Details of the licensee's risk assessment are provided in LAR. The reported risk impacts are based on a change in the Type A containment ILRT frequency from three tests in 10 years (the test frequency under 10 CFR 5 Part 50, Appendix J, Option A), to one test in 16 years. The licensee accounted for undetected containment leaks due to steel liner corrosion using the methodology that was used by Calvert Cliffs Nuclear Power Plant to estimate the likelihood and risk implications of corrosion-induced leakage of a steel liner going undetected during the extended test interval (Constellation Nuclear letter dated March 27, 2002 (ADAMS Accession No. ML020920100)). The following conclusions can be drawn from the licensee's risk analysis associated with extending the Type A ILRT frequency:

1. The reported increase in LERF for internal events was 1.37E-07 /year for internal events only. The reported increase in LERF for combined internal and external events is 6.18E-07 /year. The risk contribution from external events includes the effects of internal fires and seismic events, as discussed in Section 3.2.1 of this SE. This change in LERF is considered to be "small" per the acceptance guidelines in RG 1.174 (i.e., change in LERF between 1E-06/year and 1E-07/year). Per RG 1.174, the acceptability of a "small" risk increase requires an assessment of baseline LERF to show that the total LERF is less than 1 E-05 per reactor year. The licensee estimated the total LERF, which includes the increase in LERF associated with the change in ILRT frequency, as 6.39E-06/year, which is below the acceptance guideline for total LERF of 1 E-05/year in RG 1.17 4 for a "small" change.

2. Given a change in Type A ILRT frequency from three tests in 10 years to one test in 16 years, the LAR reported an increase in the total population dose of 3.67 person-rem per year for internal events and external events combined, or 1 percent. In response to RAl-02, the licensee further assessed the uncertainties related to the estimate of the total population dose change due to external events. In response to RAl-02, the licensee estimated an increase in the total population dose ( due to combined internal and external events) for a change in the Type A ILRT frequency from three in 10 years to once in 16 years between 3.9 person-rem/year (or 0.57 percent) and 3.81 person-rem/year (or 0.80 percent), as further discussed in Section 3.2.2 of this SE.

The updated percent increase of total population dose provided in response to RAl-02 is below the values associated with a small increase in population dose, as provided in EPRI TR-1009325, Revision 2-A, and defined in Section 3.2.4.6 of the NRC SE 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.

3. The increase in CCFP due to change in test frequency from three in 10 years to once in 16 years is 0.92 percent. This value is below the acceptance guidelines of 1.5 percentage points for a small increase in CCFP in Section 3.2.4.6 of the NRC SER for NEI 94-01, Revision 2.

Based on the risk assessment results, the NRC staff concludes that, for Indian Point 3, the increase in LERF is small and consistent with the acceptance guidelines of RG 1.17 4, and the increase in the total population dose and the magnitude of the change in the CCFP for the proposed change are small and supportive of the proposed change. The defense-in-depth philosophy is maintained, as the independence of barriers will not be degraded as a result of the requested change, and the use of the three quantitative risk metrics collectively ensures that the balance between prevention of core damage, prevention of containment failure, and consequence mitigation is preserved. Accordingly, the second condition in Section 4.2 of the SER for EPRI TR-1009325, Revision 2, is met.

3.4.4 Leak Rate for the Large Pre-Existing Containment Leak Rate Case The third condition in Section 4.2 of the SE for EPRI TR-1009325, Revision 2, stipulates that in order to make the methodology in EPRI TR-1009325, Revision 2, acceptable, the average leak rate for the preexisting containment large leak rate accident case (i.e., accident Case 3b) used by the licensees shall be 100 La instead of 35 La.

As noted by the licensee in Section 1.3 of Enclosure 1 to the LAR, the methodology in EPRI TR-1009325, Revision 2-A, incorporates the use of 100 La as the average leak rate for the preexisting containment large leak rate accident case, and this value has been used in the Indian Point 3 plant-specific risk assessments. Accordingly, the third condition in Section'4.2 of the SER for EPRI TR-1009325, Revision 2, is met.

3.4.5 Applicability if Containment Over-Pressure is Credited for ECCS Performance The fourth condition in Section 4.2 of the SE for EPRI TR-1009325, Revision 2, stipulates that in instances where containment over-pressure is relied upon for ECCS performance, a LAR is required to be submitted. In Section 5.8 of Enclosure 1 of the LAR, the licensee states that containment overpressure is not required to obtain adequate net positive suction head.

Accordingly, the fourth condition in Section 4.2 of the SE for EPRI TR-1009325, Revision 2, is not applicable.

3.4.6 Plant-Specific Confirmatory Analysis Conclusion Based on its review of the licensee's submittals, the NRC staff concludes that the licensee has demonstrated that the guidance in Section 9.2.3.1 and Section 9.2.3.4 of NEI 94-01, Revision 2-A, has been satisfied. Therefore, the staff finds that the licensee has adequately addressed the plant-specific confirmatory analysis in NEI 94-01.

Based on the above evaluation, the NRC staff finds that the licensee has used the appropriate processes, methods, and acceptance guidelines applicable to extend the ILRT test from three in 10 years to one in 15 years. In the licensee's current request to extend the ILRT test to one in 16 years, the NRC staff finds that the additional 1-year extension has been appropriately reflected in the risk evaluations. Therefore, the NRC staff concludes that the risk for the proposed LAR for a one-time extension of the Type A containment ILRT frequency to once in 16 years for Indian Point 3 meets the applicable acceptance guidelines, and therefore, is acceptable.

3.5 Technical Evaluation Summary The NRC staff finds that the licensee is satisfactorily monitoring and managing the Indian Point 3 containment and performing supplemental inspections to periodically examine and monitor aging degradation, thereby providing reasonable assurance that the containment structural and leaktight integrity will continue to be maintained. The licensee'justified the proposed change to extend the performance-based Type A ILRT test interval by demonstrating adequate performance of the containment based on plant-specific Type A ILRT test program results, consistent with the guidance in NEI 94-01.

The licensee also demonstrated satisfactory containment inspection results consistent with the inservice inspection program requirements of ASME Section XI, Subsections IWE and IWL, including as part of the license renewal application, a one-time inspection of the containment liner in 2015 to satisfy a license renewal commitment. Based on the review, the staff finds the requested one-time extension for the Type A ILRT leakage rate test frequency from 15 years to 16 years acceptable.

The LAR provided justification of the proposed one-time change in test interval described as being from a maximum of 15 years to 16 years. Primary containment leakage testing intervals have been maintained calendar-based due to variability of refueling cycle lengths, with the expectation that an ILRT would be scheduled for performance during the RFO before the interval would be exceeded. The proposed wording change to Indian Point 3 TS 5.5.15 is to an identified plant restart from RFO 3R21 and not a calendar interval.

As noted in Section 1.0 of this SE input, NEI 94-01 Revision 2-A, Section 3.1.1.2, includes a grace period not to exceed 9 months, as constrained by the requirements of the TR. As noted before, to enter into or beyond this grace period"... the licensee will have to demonstrate to the NRC staff that an unforeseen emergent condition exists." The guidance assumes a 9-month extension to the 15-year ILRT interval to be generally justifiable on the basis of an unforeseen emergent condition existing. The staff determined that the historical performance of Indian Point 3 containment regarding leakage potential suggests that the additional risk associated with a one-time nominal 12 months would be low and avoids the considerable cost and personnel radiation doses associated with performing an ILRT in order to operate the reactor for just an additional 12 months. Accordingly, the staff concludes that a Type A test interval extension from a maximum of 15 years to approximately 16 years is justified.

In summary, the staff finds that the licensee adequately implemented its performance-based Containment Leakage Rate Testing Program. The results of past ILRT and LLRT totals shown in the LAR demonstrate acceptable performance, and further demonstrate that the structural and leaktight integrity of the containment is being adequately maintained. The staff concludes that, by granting a one-time extension of the current Type A test interval requiring completion of an ILRT prior to the plant restart from RFO 3R21, there is reasonable assurance that the structural and leaktight integrity for the Indian Point 3 containment will continue to be maintained without undue risk to public health and safety. Therefore, the staff finds that, the proposed change to implement the requested one-time extension of the ILRT interval for Indian Point 3 TS 5.5.15 is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment on September 24, 2018. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (March 13, 2018; 83 FR 10916). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

Principal Contributors: L. Fields M. Biro D. Nold R. Pettis Date: December 20, 2018

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 - ISSUANCE OF AMENDMENT NO. 265 RE: ONE-TIME EXTENSION OF 10 CFR PART 50, APPENDIX J, TYPE A, INTEGRATED LEAKAGE RATE TEST INTERVAL (EPID L-2017-LLA-0406) DATED DECEMBER 20, 2018 DISTRIBUTION:

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