ML15028A308

Issuance of Amendment Extension of the Type a Containment Integrated Leak Rate Test Frequency from 10 to 15 Years

ML15028A308
Person / Time
Site:	Indian Point
Issue date:	03/13/2015
From:	Pickett D Plant Licensing Branch 1
To:	Entergy Nuclear Operations
Pickett D
References
TAC MF3426, FOIA/PA-2016-0148
Download: ML15028A308 (28)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 13, 2015 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 RE: EXTENSION OF THE TYPE A CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY FROM 10 TO 15 YEARS (TAC NO. MF3426)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 256 to Facility Operating License No. DPR-64 for the Indian Point Nuclear Generating Unit No. 3. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated February 4, 2014, as supplemented by letter dated December 9, 2014.

The amendment revises TS 5.5.15, "Containment Leakage Rate Testing Program," to allow a permanent extension of the Type A primary containment integrated leak rate test frequency from once every 10 years to once every 15 years.

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

Sincerely, Douglas V. Pickett, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-286

Enclosures:

1. Amendment No. 256 to DPR-64

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

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 FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATIONS Amendment No. 256 License No. DPR-64

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

A. The application for amendment by Entergy Nuclear Operations, Inc. (Entergy, the licensee) dated February 4, 2014, as supplemented by letter dated December 9, 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.

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 Facility Operating License No. DPR-64 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendices A, Band C, as revised through Amendment No. 256, 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.

FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: March 13, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 256 FACILITY OPERATING LICENSE NO. DPR-64 DOCKET NO. 50-286 Replace the following page of the License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 3 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 Insert Page 5.0 - 30 5.0 - 30

(4) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, Arndt. 203 to receive, possess, and use in amounts as required any 11/27/00 byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) ENO pursuant to the Act and 10 CFR Parts 30 and 70, to Arndt. 203 possess, but not separate, such byproduct and special 11/27/00 nuclear materials as may be produced by the operation of the facility.

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

(1) Maximum Power Level 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 The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 256 are hereby incorporated in the License. ENO shall operate the facility in accordance with the Technical Specifications.

(3) (DELETED) Arndt. 205 2-27-01 (4) (DELETED) Arndt. 205 2-27-01 D. (DELETED) Amdt.46 2-16-83 E. (DELETED) Amdt.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. 256

Programs and Manuals 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 exception:

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

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 are~ 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 cc/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 256

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 256 TO FACILITY OPERATING LICENSE NO. DPR-64 ENTERGY NUCLEAR INDIAN POINT 3. LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

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

1.0 INTRODUCTION

By letter dated February 4, 2014, Agencywide Documents Access and Management System (ADAMS) Accession No. ML14050A383 [Ref 1], as supplemented by letter dated December 9, 2014, ADAMS Accession No. ML14365A038 [Ref 2], Entergy Nuclear Operations, Inc. (ENO, the licensee) submitted a request for changes to the Indian Point Nuclear Generating Unit No. 3 (IP3) Technical Specifications (TSs). The proposed changes would modify TS 5.5.15 to permanently extend the frequency of the primary containment integrated leak rate test (ILRT), or Type A test, at IP3. The proposed change would extend the ILRT frequency from 10 years to 15 years on a permanent basis.

The supplemental letter dated December 9, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on July 8, 2014 (79 FR 38587) [Ref 3].

2.0 REGULATORY EVALUATION

The following explains the applicability of General Design Criteria (GDC) for IP3. The construction permit for IP3 was issued by the Atomic Energy Commission (AEC) on August 13, 1969, and the operating license was issued on December 12, 1975. The plant GDC are discussed in the Updated Final Safety Analysis Report (UFSAR) Chapter 1.3, "General Design Criteria," with more details given in the applicable UFSAR sections [Ref 4]. The AEC published the final rule that added Title 10 of the Code of Federal Regulations ( 10 CFR), Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," in the Federal Register (36 FR 3255) on February 20, 1971 [Ref 5], with the rule effective on May 21, 1971. In accordance with the U.S. Nuclear Regulatory Commission staff requirements memorandum from S. J. Chilk to J.M. Taylor, "SECY-92-223 - Resolution of Deviations Identified During the Systematic Enclosure

Evaluation Program," dated September 18, 1992 (ADAMS Accession No. ML003763736)

[Ref 6], the Commission decided not to apply the Appendix A GDC to plants with construction permits issued prior to May 21, 1971. Therefore, the GDC which constitute the licensing bases for IP3 are those in the UFSAR.

As discussed in the UFSAR, the licensee for IP3 has made some changes to the facility over the life of the unit that committed to some of the GDCs from 10 CFR Part 50, Appendix A. The extent to which the Appendix A GDC have been invoked can be found in specific sections of the UFSAR and in other IP3 licensing basis documentation, such as license amendments.

Additional documents that the NRC based its review upon included the following:

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

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

Section 50.54(0) [Ref 7] requires that primary reactor containments for water cooled power reactors be subject to the requirements set forth in Appendix J (Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors) to 10 CFR, Part 50.

Part 50 of 10 CFR, Appendix J, Option B, "Performance Based Requirements," [Ref 8] requires that a Type A test be conducted at a periodic interval based on historical performance of the overall containment system. IP3 TS 5.5.15 requires that leakage rate testing be performed as required by 10 CFR 50, Appendix J, Option B, as modified by approved exemptions, and in accordance with the guidelines contained in NRC Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Rate Testing program," dated September 1995 (ADAMS Accession No. ML003740058) [Ref 9]. This RG endorses, with certain exemptions, Nuclear Energy Institute (NEI) Report 94-01, Revision (Rev.) 0, "Industry Guideline for Implementing Performance Based Option of 10 CFR Part 50, Appendix J," dated July 26, 1995. [Ref 10]

Section 9.2.3.1, "General Requirements for ILRT Interval Extensions beyond Ten Years," of NEI 94-01, Rev. 2-A, ADAMS Accession No. ML100620847, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," [Ref 11] 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, Rev. 2-A, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals" (ADAMS Accession No. ML14024A045) [Ref 12]. 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 a safety evaluation (SE) dated June 25, 2008 (ADAMS Accession No. ML081140105)

[Ref 13], the NRC staff found the methodology in NEI 94-01, Rev. 2, and EPRI TR-1009325, Rev. 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 SE for EPRI TR-1009325, Rev. 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," [Ref 14] 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 SE for EPRI TR-1009325, Rev. 2.

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 is consistent with the clarification provided in Section 3.2.4.6 of the SE for EPRI TR-1009325, Rev. 2.

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

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

In addition, per Section 3.2.4.6 of the SE for EPRI TR-1009325, Rev. 2, the change in core damage frequency (CDF) should be calculated and reported.

References are made throughout this SE to EPRI TR-1009325, Rev. 2, NEI 94-01, Rev 2, and NEI 94-01, Rev 3. EPRI TR-1009325, Rev. 2, provides an acceptable methodology regarding PRA. NEI 94-01, Rev 2, focuses on the ILRT extension from 10 years to 15 years whereas NEI 94-01, Rev 3, focuses on extending the Type C (valve) test interval extension from 60 months to 75 months. The NRC approved each of these documents for referencing in licensing submittals pending adherence to any identified limitations and conditions. Reference 13 provides NRC approval for EPRI TR-1009325, Rev. 2, and NEI 94-01, Rev 2, and Reference 20 provides NRC approval for NEI 94-01, Rev 3. Consistent with NRC policy on approving topical reports for use in referencing in licensing applications, the industry typically resubmits the topical report with the suffix "-A" denoting that the document has been approved by the NRC. As a result, EPRI TR-1009325, Rev. 2-A, was submitted in Reference 12, NEI 94-01, Rev 2-A was submitted in Reference 11, and NEI 94-01, Rev 3-A was submitted in Reference 15. Therefore, discussions regarding these three documents in this SE referencing Revisions 2 and 2-A or 3 and 3-A should be considered synonymous.

3.0 TECHNICAL EVALUATION

3.1 PROBABILISTIC RISK ASSESSMENT 3.1.2 Plant-Specific Risk Evaluation The licensee performed a risk impact assessment for extending the Type A containment ILRT interval to 15 years. The risk assessment was provided in Attachment 3 of the LAR submitted on February 4, 2014.

In Section 4.5.1 of Attachment 1 to the LAR, the licensee stated that the plant-specific risk assessment follows the guidance in NEI 94-01, Rev. 3-A (ADAMS Accession No. ML12221A202) [Ref 15], the methodology described in EPRI TR-1009325, Rev. 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"

[Ref 18].

The licensee addressed each of the four conditions for the use of EPRI TR-1009325, Rev. 2, which are listed in Section 4.2 of the NRC SE [Ref 13]. A summary of how each condition has been met is provided in the following sections.

3.1.3 Technical Adequacy of the PRA The first condition listed in the NRC's SE stipulates that the licensee submits documentation indicating that the technical adequacy of its 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," [Ref 16] the NRC staff will use Rev. 2 of RG 1.200 to assess technical adequacy of the PRA used to support risk-informed applications received after March 2010. In Section 3.2.4.1 of the SE for NEI 94-01, Rev. 2, and EPRI TR-1009325, Rev. 2, the NRC staff states that Capability Category I of the American Society of Mechanical Engineers (ASME) PRA standard shall be applied as the standard for assessing PRA quality for ILRT extension requests, since approximate values of CDF and large early release frequency (LERF) and their distribution among release categories are sufficient to support the evaluation of changes to ILRT frequencies.

Section 4.5.2 of Attachment 1 to the LAR states that the risk assessment to support the extension of the ILRT frequency for IP3 "is based on the current Level 1 and Level 2 PRA model of record, which was released in November 2012." Appendix A to Attachment 3 of the LAR summarizes how the licensee maintains and updates the PRA model in accordance with Entergy procedures. Appendix A also states that the IP3 internal events PRA model peer review was performed against the ASME/American Nuclear Society (ANS) PRA Standard, ASME/ANS RA-Sa-2009, [Ref 17] and RG 1.200, Rev. 2, in December 2010.

The peer review indicated that 97 percent of the applicable Supporting Requirements (SRs)

"were satisfied at Capability Category II criteria or greater" for the IP3 internal events PRA model. Table A.2-2 of Appendix A of the LAR provides a summary of the 11 peer review facts

and observations (F&Os) which were classified as findings. Many of the findings identified are considered by the licensee to pertain to documentation issues or enhancements to the PRA model that do not significantly affect CDF or LERF. In all instances, the licensee determined that there would be no impact or minimal impact on CDF and LERF and that the findings would not impact the acceptability of the ILRT interval extension risk results. The NRC staff reviewed the 11 peer review F&Os to determine their impact on this application.

Finding 1-15 for SR IE-C5 states that the "initiating event frequencies are not weighted by the fraction of time the plant is at power." The licensee's disposition states that this finding remains open. In the licensee's assessment of the impact of the F&O on the ILRT extension application, the licensee notes that the approach will provide slightly conservative results. The NRC finds the licensee disposition to be acceptable because the licensee's method provides more conservative risk results by not applying a weighting factor based on plant availability.

Finding 6-6 for SR IFSO-A4 indicates that the SR is not met and states that "no distinction was made between the various causes of floods because the rupture frequencies used included all floods." The licensee's assessment of the impact of the F&O on the ILRT extension application states that "plant specific condition reports were reviewed for applicable events involving human induced flooding events, which were the only events not covered by the EPRI data." The licensee's assessment also states that no events were found and that the "documentation has been modified to specifically discuss both failure mechanisms and the conclusions of these human induced failure evaluations." The NRC finds the licensee's disposition to be acceptable.

The NRC staff reviewed the remaining 9 F&Os and the licensee's disposition and description of the impact and concluded that the F&Os have either been adequately addressed, do not have a significant impact on risk evaluations for this application, or are only documentation issues that do not affect the application.

In Section 5.7 of Attachment 3 to the LAR, the licensee performed an analysis of the external events contribution to risk and assessed the impact on the ILRT extension application. The licensee stated that the IP3 Individual Plant Examination of External Events (IPEEE) considered and screened external events such as high winds, external floods, and other hazards.

Therefore, the licensee stated that hazards other than internal fires and seismic events "are considered negligible in estimation of the external events impact on the ILRT extension assessment." Section 5.7 of Attachment 3 also states that "the method chosen to account for external events contributions is similar to that used in the Severe Accident Mitigation Alternatives analysis in which a multiplier was applied to the internal events results based on information from the IPEEE." In Section 7.0 of Attachment 3 the licensee notes that this provides a bounding assessment for the risk associated with external events.

In Section 3.2.4.2 of the SE for NEI 94-01, Rev. 2 and EPRI TR-1009325, Rev. 2, the NRC staff stated that:

Although the emphasis of the quantitative evaluation is on the risk impact from internal events, the guidance in EPRI Report No. 1009325, Rev. 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, Rev. 2)], the quality or detail will be increased or a suitable estimate of the risk impact from the external events should be performed. This assessment can be taken from existing, previously submitted and approved analyses or other alternate method of assessing an order of magnitude estimate for contribution of the external event to the impact of the changed interval."

The NRC staff concludes that the information used to estimate the effect on total LERF due to external events is acceptable. The risk impact of external events is appropriately included in the LAR and the increase in LERF was determined to meet the guidelines in RG 1.174 as discussed in Section 3.1.4 of this safety evaluation (SE).

Given that the licensee has evaluated its internal events PRA against the currently endorsed ASME PRA standard (ASME/ANS RA-Sa-2009) and the currently implemented version of RG 1.200, Rev. 2, evaluated the findings developed during the peer review of its internal events PRA for applicability to the ILRT interval extension, addressed the findings or explained their impact, and included a quantitative assessment of the contribution of external events, the NRC staff concludes that the PRA used by the licensee is of sufficient technical adequacy to support the evaluation of changes to ILRT frequency. Accordingly, the first condition (i.e., the technical adequacy of the PRA) is met.

3.1.4 Estimated Risk Increase The second condition listed in the NRC's SE 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 is consistent with the guidance in RG 1.174 and the clarification provided in Section 3.2.4.6 of the SE for NEI 94-01, Rev. 2, and EPRI TR-1009325, Rev. 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 [roentgen equivalent man] per year or 1 percent of the total population dose, whichever is less restrictive. In addition, the guidance indicates that 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 1.174. As discussed further in Section 3.1.6 of this SE, IP3 does not credit containment over-pressure; 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 4.5.3 of to the LAR. Details of the risk assessment are provided in Attachment 3. 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 are drawn from the licensee's analysis associated with extending the Type A ILRT frequency:

1. The reported increase in LERF is 1.26E-07/year for internal events only. The reported increase in LERF is 5.70E-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 3.1.3 of this SE. This change in internal and external events risk 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, an assessment of baseline LERF is required to show that the total LERF is less than 1E-05 per reactor year. The total base LERF, which includes the increase in LERF associated with the change in ILRT frequency, is estimated to be 6.35E-06/year which is below the total LERF value of 1E-05 per reactor year in RG 1.174.

2. Per Table 5.7-3 in Attachment 3 to the LAR, given a change in Type A ILRT frequency from three tests in 1O years to one test in 15 years, the reported increase in the total population dose is 3.38 person-rem per year for internal events and external events combined, or 0.93 percent of the total population dose. The percent increase of total population dose is below the values associated with a small increase in population dose, as provided in EPRI TR-1009325, Rev. 2-A, and defined in Section 3.2.4.6 of the NRC SE for NEI 94-01, Rev. 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 licensee reports a 0.85 percent increase in CCFP for going from a test frequency of three tests in 10 years to one test in 15 years. This is below the guideline value of 1.5 percentage points for a small increase in CCFP, as provided in EPRI TR-1009325, Rev. 2-A, and defined in Section 3.2.4.6 of the NRC SE for NEI 94-01, Rev. 2.

Therefore, the NRC staff concludes that the increase in LERF is small and consistent with the acceptance guidelines of RG 1.174. In addition, the staff finds the increase in the total integrated plant risk and the magnitude of the change in the CCFP for the proposed change are also small and supportive of the proposed change. Therefore, 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 (i.e., the estimated risk frequency) is met.

3.1.5 Leak Rate for the Large Pre-Existing Containment Leak Rate Case The third condition listed in the NRC's SE stipulates that in order to make the methodology in EPRI TR-1009325, Rev. 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 Table 4.5-1 in Section 4.5.1 of Attachment 1 to the LAR, and in the footnote for Section 1.3 in Attachment 3 to the LAR, this value has been used in the IP3 analysis. Accordingly, the third condition is met.

3.1.6 Applicability if Containment Over-Pressure is Credited for ECCS Performance The fourth condition listed in NRC's SE stipulates that in instances where containment over-pressure is relied upon for ECCS performance, a LAR is required to be submitted. As noted by

the licensee in Table 4.5-1 in Section 4.5.1 of Attachment 1 to the LAR, IP3 does not rely on containment over-pressure for ECCS performance. Accordingly, the fourth condition is met.

3.2 DETERMINISTIC CONSIDERATION NEI 94-01, Rev. 0, specifies an initial test interval of 48 months, but allows an extended interval of 1O 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 IP3 was approved by letter dated April 17, 2001 (ADAMS Accession No. ML011070447) [Ref 19]. 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 IP3 were successful, so the current interval requirement would be 10 years per current TS 5.5.14 and is required to be performed no later than March 2015. The proposed amendment would allow the next ILRT for IP3 to be performed within 15 years from the last ILRT (i.e., March 2005), therefore the next Type A test would be performed on or before March 2020 as opposed to the currently scheduled ILRT date of March 2015.

In the NRC's SE dated June 25, 2008 [Ref 13], the staff concluded that NEI 94-01, Rev. 2 was 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 with regards to containment leakage rate testing, subject to the limitations and conditions noted in Section 4.1 of the SE. 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."

By letter dated June 8, 2012 (ADAMS Accession No. ML121030286) [Ref 20], the NRC published an SE, with limitations and conditions, for NEI 94-01, Rev. 3. In the SE, the NRC concluded that NEI 94-01, Rev. 3 is acceptable for referencing by licensees proposing to amend their TS regarding Option Band C testing. Limitations and conditions are stated in Section 4.0 of the SE:

Condition 1 The approved NEI 94-01 is allowing Type C LLRTs [Local Leak Rate Tests] to be increased to 75 months with the permissible extension to 84 months for non-routine emergent conditions. This is subject to certain exceptions.

Condition 2 The basis for acceptability of extending the ILRT interval out to once per 15 years was the enhanced and robust primary containment inspection program and the local leakage rate testing of penetrations. For the purposes of assessing and monitoring or trending overall containment leakage potential, the as-found minimum pathway leakage rates for the just tested penetrations are summed with the as-left minimum pathway leakage rates for penetrations tested during the

previous one, two, or even three refueling outages. Given the required margin included with the performance criterion and the considerable extra margin most plants consistently show with their testing, any understatement of the LLRT total using a 5-year test frequency is thought to be accounted for conservatively.

Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of trending.

Both of these limitations and conditions would apply if Entergy applies for an LAR requesting an extension of the Type C leakage testing beyond 60 months. However, the licensee stated that it plans on retaining the 60-month limitation on Type C penetrations and therefore these limitations and conditions do not apply.

3.2.1 Technical Specifications TS 5.5.15, "Containment Leakage Rate Testing Program," currently states:

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 the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak Test Program, dated September 1995" as modified by the following exception:

ANS 56.8-1994, Section 3.3.1: WCCPPS isolation valves are not Type C tested.

The proposed change to TS 5.5.15 in the application of Reference 1 stated:

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, Rev. 3A, "Industry Guidelines for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," July 12, 2012, as modified by the following exception:

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

Contrary to NRC expectations when approving NEI 94-01, Rev 3 [Ref 20], the limitations and conditions previously provided in the SE for NEI 94-01 Rev. 2-A [Ref 11] were not incorporated into NEI 94-01 Rev. 3-A [Ref 15]. As a result, reference to NEI 94-01, Rev. 3-A without also referencing the limitations and conditions of NEI 94-01, Rev. 2-A would not be acceptable if a permanent Type A test interval extension to beyond 10 years were contemplated. Therefore, in a teleconference on November 12, 2014, the NRC staff advised Entergy that NEl-94-01, Rev. 2A, was the appropriate reference for guidance providing for permanent extension of the Type A testing interval to 15 years; and NEI 94-01 Rev. 3A, is appropriate for guidance providing for a permanent extension of Type C test intervals beyond 60 months in addition to

the Type A test interval of 15 years. As a result of this conversation, the licensee submitted an updated proposed TS change in its supplement dated December 9, 2014, which stated:

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, Rev. 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 leakage rate testing requirements of 10 CFR 50, Appendix J, Option B (Type A ILRT and Type B and Type C LLRTs) and the containment lnservice Inspection (ISi) requirements mandated by 10 CFR 50.55a, together, help ensure the continued leak-tight and structural integrity of the containment during its service life. The NRC staff's review of the information presented in the LAR, as supplemented by letter dated December 9, 2014, is presented in Section 3.2.2 of this SE.

3.2.2 Containment ISi Program and Containment Leak-Tight Integrity Considerations As required by 10 CFR 50.54(0), IP3 is subject to the requirements set forth in 10 CFR 50, Appendix J. Option B of Appendix J which requires that test intervals for Type A, Type B, and Type C testing be determined by using a performance-based approach. Currently, the IP3 ILRT program is based on RG 1.163, which endorses NEI 94-01, Rev. 0. The LAR proposes to revise the IP3 ILRT program by implementing the guidance in NEI 94-01, Rev. 2-A.

As previously stated, the NRC found NEI 94-01, Rev 2, acceptable for referencing by licensees subject to the limitations and conditions noted in Section 4.0 of the SE. The licensee stated that it meets the limitations and conditions of the SE. NEI 94-01, Rev. 2, which was issued in 2008, included provisions for extending the Type A ILRT interval to 15 years. NEI 94-01,Rev. 3-A, was issued in July 2012, and included guidance for extending the Type C LLRT interval to 75 months, but inadvertently omitted the limitations and conditions contained in the SE incorporated into Rev. 2-A. 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 licensee addressed the limitations and conditions contained in Section 4.1 of the staff's SE for NEI 94-01 Rev. 2-A as follows:

Condition 1 For calculating the Type A leakage rate, the licensee should use the definition in the NEI 94-01, Rev. 2 in lieu of that in ANSl/ANS-56.8-2002.

Staff Assessment of Condition 1 The licensee addressed Condition 1 in its February 4, 2014, letter. The licensee uses the definition found in Section 5.0 of NEI 94-01, Rev. 3-A for calculating the Type A leakage rate (GTP- 315 Step 4.1.4.A). NEI 94-01, Rev. 3-A contains the same definition as NEI 94-01, Rev. 2. The licensee stated that the ILRT testing history of the containment structure leakage is acceptable, with margin, however there were two incidents with as-found leakages above normal, which are discussed in Table 1 below.

Type A Tests 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 IP3 were performed in December 1990 and March 2005. The licensee stated that the value of Pa for IP3 is 42.38 psig per TS 5.5.15.

The results provided below show the two most recent tests performed in 1990 and 2005 for IP3 as well as past Type A ILRT results going back to 1978. Below is a summary table of those Type A tests, which were successful, with containment performance leakage rates less than the La (1.0 La at Pa) of 0.1 percent containment air weight per day. The NRC staff finds that consistent with the guidance in NEI 94-01, Rev. 3-A (which is similar to guidance in NEI 94-01, Rev. 2-A), this performance history for Type A tests supports extending the current ILRT interval to 15 years.

Table 1. Indian Point Type A ILRT Results Date As found Leakage Test Pressure (psia)

(% Containment weight per day)

March 2005 0.0565 60.61 December 1990 0.032 59.49 Julv 27, 1987 0.34 59.89 August4, 1982 0.034 60.00 August 2, 1978 0.14 60.00 The licensee stated in its July 27, 1987, ILRT test, "There was a leak through the reactor coolant pump seal water return valve MOV-222 on penetration R, Line 17. Valve 221A was closed to isolate MOV-222 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 past two ILRT Type A results of the containment are acceptable with reference to the TS criterion of 0.1 percent leakage of containment air weight per day (wt percent/day) at the design basis loss of coolant accident pressure (La). Since the last two IP3 Type A as-found results were less than 1.0 La, a test frequency of at least once per 15 years would be in accordance with NEI 94-01, Rev. 2-A.

Type B and C Tests The licensee provided a history of Type B and C tests in accordance with 10 CFR 50, Appendix J, Option B. The licensee stated that the minimum pathway combined Type B and Type C leakage from the March 2005 outage (during which a Type A test was performed), and are listed in the table below. The table also contains subsequent combined as-found Type B and Type C test values during each consecutive outage since the last Type A test, as provided in the table below.

Date As-Found La (ccm) Percent ((As- Percent ((As-Leakage (seem} Found/la) x 100) Found/.6La)) x 100)

April 2005 41585 119689 0.208 0.347 April 2007 30352 119689 0.152 0.254 April 2009 44621 119689 0.224 0.373 April 2011 51878 119689 0.260 0.433 April 2013 36669 119689 0.184 0.306 Based on the information provided by the licensee, the staff concludes that the combined as-found minimum pathway Type B and C leakage values are well below the allowable leakage (0.6 La) and there is no discernible adverse trend in these values during this period.

Condition 2 The licensee submitted a schedule of containment inspections to be performed prior to, and between, Type A tests.

Staff Assessment of Condition 2 The licensee addressed Condition 2 in its February 4, 2014, letter. The licensee stated that it will implement the requirement contained in NEl-94-01 3A, Section 9.2.3.2 (which is the same definition described in NEI 94-01, Rev. 2):

To provide continuing supplemental means of identifying potential containment degradation, a general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leak-tight integrity must be conducted prior to each Type A test and during at least three other outages before the next Type A test if the interval for the Type A test has been extended to 15 years. It is recommended that these inspections be performed in conjunction or coordinated with the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE/IWL required examinations.

The licensee included a table that lists current and planned containment inspections assuming the Type A test frequency is extended to 15 years. In addition, the licensee stated that IP3 has established procedures for performing visual examinations of the accessible surfaces of the containment for detection of structural problems. The licensee stated that the visual examinations performed are in accordance with its ISi program which satisfies requirements specified in Option B. Evaluations of inaccessible areas are also addressed in its program, in accordance with 10 CFR 50.55a(b)(2)(ix). The guidance in NEl-94-01, Rev. 3-A, Section

9.2.3.2, specifies that these examinations should be conducted prior to initiating a Type A test and during three other outages before the next Type A test if the interval for the Type A test has been extended to 15 years in order to allow for early detection of evidence of structural deterioration. These visual examinations have been completed, and the licensee discussed corrective actions identified by these inspections in its LAR.

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

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, Rev. 3-A, and NEI 94-01, Rev. 2-A.

Condition 3 The licensee addressed the areas of the containment structure potentially subjected to degradation.

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

Staff Assessment of Conditions 3 & 4 In its application dated February 4, 2014, the licensee documented deficiencies in past IWE and IWL inspections. However, the licensee stated that all of the conditions observed were considered minor or they were further evaluated and noted to be acceptable. 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 IP3 IWE/IWL Containment ISi Plans which implement the requirements of the ASME,Section XI, Subsections IWE and IWL, as required by 10 CFR 50.55a(g).

The licensee provided a summary of LLRT results which demonstrated acceptable performance history in accordance with the Containment Leakage Rate Program. The NRC staff finds that the licensee has effectively addressed Condition 3.

In its letter of February 4, 2014, the licensee stated that any tests and inspections performed following major modifications to the containment structure will be in accordance with the ASME Containment In-Service Inspection Program, ASME Appendix J (Primary Containment Leak Rate Testing) Program, and ASME Section XI.

The NRC staff finds that the licensee has effectively addressed Condition 4 by stating that tests and inspections performed following major modifications to the containment structure will be in accordance with the ASME Containment In-Service Inspection Program, ASME Appendix J (Primary Containment Leak Rate Testing) Program, and ASME Section XI.

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, Rev. 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.

Staff Assessment of Condition 5 The licensee stated that it plans on using the Type A interval definition contained in NEI 94-01, Rev. 2-A and NEI 94-01, Rev. 3-A. The licensee stated that it is not pursuing an extension to the ILRT interval for more than 15 years. Therefore, Condition 5 is not currently applicable.

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, Rev. 2, and EPRI Report No. 1009325, Rev. 2, including the use of past containment ILRT data.

Staff Assessment of Condition 6 Condition 6 is only applicable to plants licensed under 10 CFR Part 52. IP3 is not licensed pursuant to 10 CFR Part 52 and this condition is not applicable.

SUMMARY

Based on the evaluation above, the NRC staff finds that the licensee has effectively addressed all applicable Limitations and Conditions of Section 4.1 for NEI 94-01, Rev 2, and Section 4.2 for EPRI TR-1009325, Rev 2, of the staff's SE (Ref 13). The licensee has identified the areas of containment potentially subject to degradation, and committed to conduct tests or inspections following major modifications to the containment. Further, the licensee has implemented an adequate Containment Leakage Rate Testing ILRT and LLRT program, and supplemental inspections to periodically examine, monitor, and manage age-related and environmental degradation of the IP3 primary containment. As a result, the staff finds that the licensee's proposal to permanently extend the frequency of the Type A ILRT from 10 years to 15 years 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. By letter dated February 20, 2015 [Ref 21 ], the State official submitted comments recommending that the NRC staff deny the proposed license amendment.

The letter from the State of New York described the importance of the ILRT and stated, in part,

that this is the only test that could identify "containment leakage past structural elements that cannot be locally leak rate tested such as containment liner seal welds." The letter noted that the last ILRT conducted at IP3 was in 2005 and if the NRC grants the request to extend the ILRT frequency to once every 15 years and if license renewal is granted, no ILRT would be conducted prior to license renewal, and only one more ILRT would be performed during the continued operating life of IP3.

The State of New York characterized Entergy's justification for the proposed license amendment as "legally impermissible" because the licensee included costs of the ILRT, radiation exposures, and extensions of critical path time as part of their overall basis for the request. The State noted that the NRC may not consider costs when a requirement is part of the "adequate protection" licensing basis.

While the State of New York acknowledged that Entergy's request references industry topical reports that have been reviewed and approved by the NRC, the State believes that Entergy's risk assessment fails to adequately account for several factors that are unique to IP3. These factors include IP3's high core damage frequency relative to plants in the central and eastern United States, seismic considerations that are currently under review by the NRC staff as part of the Fukushima related activities, potential sabotage, the presence of major airports in the New York City vicinity, the large permanent and transient population that is present in the New York City vicinity, the proximity of the IP3 site to reservoirs that provide drinking water to the city, along with other unique site aspects that include transportation links, financial, scientific, medical, educational, and historical resources.

In 1995, the NRC amended its regulations to provide a performance-based option for leakage-rate testing of containments of light-water-cooled nuclear power plants. (Final Rule, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, 60 Fed. Reg. 49,495, Sept. 26, 1995). The performance-based "Option B" is available for voluntary adoption by licensees in lieu of compliance with the "Option A" prescriptive requirements in the Commission's regulations. When promulgating the rule, the Commission observed that the new rule improves the focus of the regulations by eliminating prescriptive requirements that are marginal to safety. Further, the Commission noted that the final rule allows test intervals to be based on system and component performance and provides licensees greater flexibility for cost-effective implementation methods of regulatory safety objectives.

In its comments, the State of New York asserts that Entergy's justification for its LAR is "legally impermissible" because Entergy discussed, in part, significant savings in cost during future refueling outages. As explained below, when the Commission promulgated the rules allowing licensees to request performance-based containment leak rate testing, the Commission explicitly described the cost savings potential. Accordingly, Entergy's discussion about cost-saving does not amount to an impermissible request, but instead is consistent with what is expected under the rules.

The rule was part of the Commission's "Regulatory Improvement Program" whereby the Commission embraced a performance-based regulatory approach based on the following framework:

1. The new performance-based regulation will be less prescriptive and will allow licensees the flexibility to adopt cost-effective methods for implementing the safety objectives of the original rule.

2. The regulatory safety objectives will be derived, to the extent feasible and practical, from risk considerations with appropriate consideration of uncertainties, and will be consistent with the NRC's Safety Goals.

3. Detailed technical methods for measuring or judging the acceptability of a licensee's performance relative to the regulatory safety objectives will be, to the extent practical, provided in industry standards and guidance documents which are endorsed in NRC regulatory guides.

4. The new regulation will be optional for current licensees so that licensees can decide to remain in compliance with current regulations.

5. The regulation will be supported by necessary modifications to, or development of, the full body of regulatory practice including, for example, standard review plans, inspection procedures, guides, and other regulatory documents.

6. The new regulation will be formulated to provide incentives for innovations leading to improvements in safety through better design, construction, operating, or maintenance practices.

Concerning costs, when making the rule in 1995, the Commission stated at 60 Fed. Reg.

49,498: "Estimates of the remaining industry-wide costs of implementing current Appendix J requirements ranged from $720 to $1,080 million, approximately 75 percent of which could be averted with a performance-based rule."

Accordingly, it is the Commission's expectations that licensees will save money by utilizing the performance-based aspects of the Commission's rules, and Entergy's statement that it expects to save money is congruent with the Commission's expectations. It is not legally impermissible.

However, although cost-saving is expected under the Commission's rules, cost savings is not a standard used to determine if an amended license should be issued, thus the State of New York's concern, which is that cost-saving will be considered, should be assuaged. Instead, whenever a holder of an operating license desires to amend the license, an application for an amendment must be filed with the Commission fully describing the changes desired, and following as far as applicable, the form prescribed for original applications. In determining whether an amendment to a license will be issued to the applicant, the Commission will be guided by the considerations which govern the issuance of initial licenses to the extent applicable and appropriate. For an amendment to an operating license, and as described in 10 CFR 50.57(a), the considerations include findings that 1) the facility will operate in conformity with the application as amended, the provisions of the Atomic Energy Act, and the rules and

regulations of the Commission; 2) there is reasonable assurance that (i) the activities authorized by the operating license can be conducted without endangering the health and safety of the public, and (ii) such activities will be conducted in compliance with the regulations in this chapter; and 3) the issuance of the license amendment will not be inimical to the common defense and security or to the health and safety of the public.

One of the conditions required of all operating licenses for light-water-cooled power reactors as specified in 10 CFR § 50.54(0) is that primary reactor containments meet the leakage-rate test requirements in either Option A (Prescriptive Requirements) or Option B (Performance-Based Requirements) of Appendix J to 10 CFR Part 50. These test requirements ensure that (a) leakage through these containments or systems and components penetrating these containments does not exceed allowable leakage rates specified in the technical specifications; and (b) integrity of the containment structure is maintained during its service life.

Pursuant to 10 CFR § 50.54(0), any amended license must continue to meet the leakage-rate test requirements in either Option A or Option B of Appendix J to 10 CFR Part 50. As allowed by, and specified in, Option B of Appendix J to 10 CFR part 50, Type A tests must be conducted at a periodic interval based on the historical performance of the overall containment system as a barrier to fission product releases to reduce the risk from reactor accidents.

Significantly, while the NRC will factor in cost where such consideration is mandated (e.g., 10 CFR 50.109 (backfitting) considers, in some circumstances, whether direct and indirect costs of implementation are justified in view of increased protection), cost is not a factor to be considered under 10 CFR § 50.57(a), 10 CFR § 50.54(0), and leakage-rate requirements in Appendix J to 10 CFR part 50.

Consistent with the above, the NRC staff did not consider cost-savings when reviewing Entergy's LAR.

The State of New York also commented that the testing interval "has been continuously degraded" though the extended leakage test intervals. The increase in testing intervals was anticipated by the Commission, as discussed in its rulemaking. Specifically, the Commission noted: "With respect to the 10-year interval for ILRTs, the NRC believes its technical support document (NUREG-1493 [Performance-Based Containment Leak Test Program]) is persuasive by demonstrating that testing intervals could be increased up to once every 20 years with an imperceptible increase in risk, using actual ILRT data which accounted for random and plant-specific failures and plant aging effects." 60 Fed. Reg. at 49,498.

The State of New York also expressed concern over testing frequency and interval by highlighting the limited number of tests that would occur during the period of extended operation under a renewed operating license. As discussed in Section 3 of this SE, Entergy provided sufficient information for the NRC staff to find that the new periodic interval is sufficient. To this end, the staff found that a "Type A" test after every fifteen years of operation is sufficient.

The State of New York further commented that unique factors relating to the site had not been considered. As previously discussed, the Commission will be guided by the considerations which govern the issuance of initial licenses to the extent applicable and appropriate. The State of New York highlights a broad range of items, including sabotage, earthquakes, core damage

frequency, and the local population. The NRC staff considered the state's comments, and the relationship between the comments and Entergy's request concerning changing the testing frequency under its performance-based program. However, the comments raised by the state did not provide sufficient information to show that Energy's application was deficient.

Lastly, the State of New York asserted that Entergy's risk assessment did not properly account for the fact that Entergy has requested a renewed license for Indian Point Unit 3, and that "Entergy's risk assessment time intervals do not match the reality of Indian Point Unit 3 circumstances." The State says that "Unit 3 will operate 30 years with only one Type A containment test." In fact, under the proposed amendment, the 30-year period would be bookended by leak-rate tests, thus the maximum period between tests would be 15 years. This period is consistent with Entergy's application.

In summary, after careful consideration of the comments provided by the State of New York, the NRC staff did not identify any deficiency in Entergy's application based on the information provided by the state.

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 (79 FR 38587). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

7.0 REFERENCES

1. Ventosa, John A., Site Vice President, Entergy Nuclear Operations, Inc. (ENO), letter to Document Control Desk, U.S. Nuclear Regulatory Commission (NRC), "Proposed License Amendment Regarding Extending the Containment Type A Leak Rate Testing Frequency to 15 years," NL-14-014, for Indian Point Unit Number 3, February 4, 2014, Agencywide Documents Access and Management System (ADAMS) Accession No. ML14050A383.

2. Ventosa, John A., Site Vice President, Entergy Nuclear Operations, Inc. (ENO), letter to Document Control Desk, U.S. Nuclear Regulatory Commission (NRC), "Revision to Proposed License Amendment Regarding Extending the Containment Type A Leak Rate Testing Frequency to 15 years," NL-14-144, for Indian Point Unit Number 3, December 9, 2014, ADAMS Accession No. ML14365A038.

3. NRC, "Biweekly Notice, Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations," Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York, Vol. 79, No. 130, July 8, 2014, pp. 38587-38588 (79 FR 38587).

4. NRC, "Updated Final Safety Analysis Report (UFSAR), - Chapter 1.3, General Design Criteria", [Redacted Version] for Indian Point Nuclear Generating Unit No. 3, June 2011, ADAMS Accession No. ML11236A228.

5. Federal Regulation U.S. Code of Federal Regulations, "Domestic Licensing of Production and Utilization Faciliies - General Design Criteria for Nuclear Power Plants,"

Title 10, of the Code of Federal Regulations (10 CFR) Part 50, Appendix A."

6. Chilk, Samuel J. Secretary, Staff Requirements Memo to James M. Taylor, Executive Director of Operations, "SECY-92-223 - Resolution of Deviations Identified During The Systematic Evaluation Program," September 18, 1992, ADAMS Accession No. ML003763736.

7. Federal Regulation U.S. Code of Federal Regulations, "Domestic Licensing of Production and Utilization Facilities - Systems Penetrating Containment," Title 1O, of the Code of Federal Regulations (10 CFR) Part 50, Section 50.54(0).

8. Federal Regulation U.S. Code of Federal Regulations, "Domestic Licensing of Production and Utilization Faciliies - Primary Reactor Containment Leakage Testing For Water-Cooled Power Reactors - Performance Based Requirements," Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix J, Option B.

9. U.S. Nuclear Regulatory Commission (NRC), Regulatory Guide 1.163, "Performance Based Containment Leak-Rate Testing Program," September 1995, ADAMS Accession No. ML003740058.

10. Butler, John C., Director, Safety Focused Regulation, Nuclear Generation Division, Nuclear Energy Institute, "NE/ 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, "July 26, 1995.

11. Butler, John C., Director, Safety Focused Regulation, Nuclear Generation Division, Nuclear Energy Institute, "NE/ 94-01 Revision 2-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, November 19, 2008, ADAMS Accession No. ML100620847.

12. Electric Power Research Institute (EPRI) Technical Report (TR), "Risk Impact of Extended Integrated Leak Rate Testing Intervals," Revision 2-A of 1009325, ADAMS Accession No. 14024A045.

13. Maxin, Mark J. , Acting Deputy Director, NRC, Letter to Mr. John C. Butler, Director, Safety Focused Regulation, Nuclear Generation Division of Nuclear Energy Institute, "Final Safety Evaluation for Nuclear Energy Institute (NE/) 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 Testing Intervals," June 25, 2008, ADAMS Accession No. ML081140105.

14. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200 "Regulatory Guide 1.200 Implementation," dated March 2009, ADAMS Accession No. ML090410014.

15. Bradley, Biff, Director, Risk Assessment, Nuclear Energy Institute (NEI)

Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, NE/ 94-01 Revision 3-A, June 9, 2011, ADAMS Accession No. ML12221A202.

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17. American Society of Mechanical Engineers (ASME), American Nuclear Society (ANS)

"Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," PRA Standard - RA-Sa-2009." February 2, 2009, New York, NY.

18. U.S. Nuclear Regulatory Commission (NRC), Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes To The Licensing Basis," July 1998, ADAMS Accession No. ML003740133.

19. Wunder, George F., Project Manager, NRC, Letter to Mr. Michael Kansler, Vice President and Chief Operating Office, Entergy Nuclear Operatins, Inc., "Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: Frequency of

Performance-Based Leakage Rate Testing (TAC NO. MB0178)" April 17, 2001, ADAMS Accession No. ML011020315, Package No. ML011070447.

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Report, 94-01, Revision 3, 'Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J,' June 8, 2012, ADAMS Accession No. ML121030286.

21. Comments submitted by the New York State Energy Research and Development Authority, February 2015, ADAMS Accession No. ML15055A512.

Principal Contributors: David Gennardo Samina Shaikh Date: March 13, 2015

ML15028A308 OFFICE LPL 1-1/PM LPL 1-1/LA APLA/BC STSB/BC NAME DRender KGoldstein HHamzehee by memo dated RElliott DATE 02/01/2015 02/01/2015 12/17/2014 02/ 04 /2015 OFFICE SC VB/BC OGC LPL 1-1/PM LPL 1-1/BC NAME RDennig by memo dated STurk NLO DPickett BBeasley w/changes DATE 01/08/2015 02/06/2015 02/03/2015 03 /13/2015