ML16351A460

Issuance of Amendment to Extend Containment Leakage Test Frequency

ML16351A460
Person / Time
Site:	Fermi
Issue date:	03/09/2017
From:	Sujata Goetz Plant Licensing Branch III
To:	Fessler P DTE Electric Company
Goetz S, NRR/DORL/LPLIII, 415-8004
References
CAC MF7534
Download: ML16351A460 (44)

Text

{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 9, 2017 Mr. Paul Fessler Senior Vice President and Chief Nuclear Officer DTE Electric Company Fermi 2 - 210 NOC 6400 North Dixie Highway Newport, Ml 48166

SUBJECT:

FERMI 2 NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT TO EXTEND CONTAINMENT LEAKAGE TEST FREQUENCY (CAC NO. MF7534)

Dear Mr. Fessler:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment No. 205 to Renewed Facility Operating License (RFOL) No. NPF-43 for Fermi 2 nuclear power plant. The amendment is in response to your application dated March 22, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16082A309), as supplemented by letter dated August 11, 2016 (ADAMS Accession No. ML16225A493). The amendment revises Technical Specification (TS) 5.5.12, "Primary Containment Leakage Rate Testing Program," for the permanent extension of the Type A test interval up to one test in 15 years, as stipulated in Nuclear Energy Institute (NEI) 94-01, Revision 2-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," October 2008 (ADAMS Accession No. ML100620847). The amendment also increases the containment isolation valves leakage test intervals (i.e., Type C tests) from their current 60 months to 75 months by replacing TS 5.5.12.a. reference to Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program" (ADAMS Accession No. ML003740058), with a reference to NEI 94-01, Revision 3-A (ADAMS Accession No. ML12221A202), and the conditions and limitations specified in NEI 94-01, Revision 2-A, to implement the performance-based leakage testing program in accordance with Title 10 of the Code of Federal Regulations Part 50, Appendix J, Option B. In addition, the amendment deletes from TS 5.5.12.a, a Type A test extension that expired in 2007 and also removes from Fermi 2 RFOL paragraph 2.D an exemption, listed as "(c)," from Appendix J testing requirements for containment air locks. Amendment No.108, approved by the NRC in 1996 (ADAMS Accession No. ML020730597), eliminated the need for this exemption by adopting Appendix J, Option B, requirements for the testing program for containment air locks.

P. Fessler A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely.

                                                &+/-~Manager Plant Licensing Branch 111 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-341

Enclosures:

1. Amendment No. 205 to NPF-43

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DTE ELECTRIC COMPANY DOCKET NO. 50-341 FERMI 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 205 Renewed License No. NPF-43

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

A. The application for amendment by the DTE Electric Company (DTE, the licensee) dated March 22, 2016, as supplemented by letter dated August 11, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Enclosure 1

2. Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No.

NPF-43 is hereby amended to read as follows: Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 205, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this renewed license. DTE Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. Accordingly, Paragraph 2.D of Renewed Facility Operating License No. NPF-43 is hereby amended by inserting "and" before "(b)" and by deleting the following text:

               ; and (c) an exemption from the requirement of Paragraph lll.D.2(b)(ii) of Appendix J, the testing of containment air locks at times when containment integrity is not required (Section 6.2. 7 of SSER #5)

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

FOR THE NUCLEAR REGULATORY COMMISSION K\Mv'7\%~yw-{/ fer

                                                           * . j 1,,/

David J. Wrona, Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-43 and Technical Specifications Date of Issuance: March 9 , 2O1 7

FERMI 2 ATTACHMENT TO LICENSE AMENDMENT NO. 205 RENEWED FACILITY OPERATING LICENSE NO. NPF-43 DOCKET NO. 50-341 Replace the following pages of the Renewed Facility Operating License and Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. REMOVE INSERT License License 4 4 8 8 TS TS 5.0-18 5.0-18

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 205, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this renewed license. DTE Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. (3) Antitrust Conditions DTE Electric Company shall abide by the agreements and interpretations between it and the Department of Justice relating to Article I, Paragraph 3 of the Electric Power Pool Agreement between DTE Electric Company and Consumers Power Company as specified in a letter from The Detroit Edison Company to the Director of Regulation, dated August 13, 1971, and the letter from Richard W. Mclaren, Assistant Attorney General, Antitrust Division, U.S. Department of Justice, to Bertram H. Schur, Associate General Counsel, Atomic Energy Commission, dated August 16, 1971. (4) Deleted (5) Deleted (6) Deleted (7) Deleted (8) Deleted (9) Modifications for Fire Protection (Section 9.5.1, SSER #5 and SSER #6)* DTE Electric Company shall implement and maintain in effect all provisions of the approved fire protection program as described in its Final Safety Analysis Report for the facility through Amendment 60 and as approved in the SE through Supplement No. 5, subject to the following provision: (a) DTE Electric Company may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

• The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report (SER) and/or its supplements wherein the license condition is discussed.

Renewed License No. NPF-43 Amendment No. 205

D. Exemptions from certain requirements of Appendices E and .J to 10 CFR Part 50, are described in supplements to the SER. These include: (a) an exemption from the requirement of Section IV.F of Appendix E that a full participation emergency planning exercise be conducted within one year before issuance of the first operating license for full power and prior to operation above five percent of rated power (Section 13.3 of SSER #6); and (b) an exemption from the requirement of Paragraph lll.C.2(b) of Appendix J, the testing of the main steam isolation valves at the peak calculated containment pressure associated with the design basis accident (Section 6.2.7 of SSER #5). These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. Therefore, these exemptions are hereby granted pursuant to 10 CFR 50.12. With the granting of these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission. E. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Fermi 2 Physical Security Plan, Security Training and Qualification Plan, and Safeguards Contingency Plan" submitted by letter dated September 9, 2004, and supplemented on October 7, 2004, and October 14, 2004, November 18, 2005, and May 18, 2006. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Fermi 2 CSP was approved by License Amendment No. 185, as supplemented by License Amendment 200. F. Deleted G. The licensees shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims. Amendment No. 485, 200 i2 0 5

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP) (continued) The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. 5.5.12 Primary Containment Leakage Rate Testing Program

a. A program shall be established to implement the leakage rate testing of the primary 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 Nuclear Energy Institute (NEI) 94-01. Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 2012, and the limitations and conditions specified in NEI 94-01, Revision 2-A, dated October 2008.

b. The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa. is 56.5 psig.

c. The maximum allowable containment leakage rate La, at Pa, shall be 0.5% of containment air weight per day.

d. Leakage Rate acceptance criteria are:

1. Containment leakage rate acceptance criterion is s 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 La for the required Type Band C tests ands 0.75 La for Type A tests.

2. Air lock testing acceptance criteria are:

i) Overall air lock leakage rate is s 0.05 La when tested at ~ Pa. ii) For each door, leakage rate is s 5 scf per hour when the gap between the door seals is pressurized to ~ Pa. (continued) FERMI - UNIT 2 5.0-18 Amendment No. lM, ' 205

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 205 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-43 DTE ELECTRIC COMPANY FERMI 2 DOCKET NO. 50-341

1.0 INTRODUCTION

By application dated March 22, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16082A309), as supplemented by letter dated August 11, 2016 (ADAMS Accession No. ML16225A493}, DTE Electric Company (the licensee) submitted a license amendment request (LAR) for Fermi, Unit 2 (Fermi 2). The LAR proposes changes to Technical Specification (TS) 5.5.12, "Primary Containment Leakage Rate Testing Program," for the permanent extension of the Type A test interval to one test in 15 years, as stipulated in Nuclear Energy Institute (NEI) 94-01, Revision 2-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," October 2008 (ADAMS Accession No. ML100620847). The LAR also proposes to increase the containment isolation valves (CIVs) leakage test intervals (i.e., Type C tests) from their current 60 months to 75 months by replacing TS 5.5.12.a reference to Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program" (ADAMS Accession No. ML003740058), with a reference to NEI 94-01, Revision 3-A (ADAMS Accession No. ML12221A202), and the conditions and limitations specified in NEI 94-01, Revision 2-A, to implement the performance-based leakage testing program in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix J, Option B. In addition, the amendment proposes to delete from TS 5.5.12.a, a Type A test extension that expired in 2007 and also proposes to remove from Fermi 2 Renewed Facility Operating License (RFOL) paragraph 2.D an exemption, listed as "(c)," from Appendix J testing requirements for containment air locks. Amendment No.108, approved by the NRC in 1996 (ADAMS Accession No. ML020730597), eliminated the need for this exemption by adopting Appendix J, Option B, requirements for the testing program for containment air locks. The August 11, 2016, supplement provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazard consideration determination as published in the Federal Register. Enclosure 2

2.0 REGULATORY EVALUATION

Paragraph (o) of 10 CFR 50.54, "Conditions of licenses, requires that primary reactor containments for water-cooled power reactors be subject to the requirements in 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." Appendix J contains two options: Option A - Prescriptive Requirements and Option B - Performance-Based Requirements, either of which can be used to meet Appendix J requirements. The testing requirements in 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. Fermi 2 has been voluntarily implementing Option B for meeting the requirements of Appendix J. Option B, Section V.B.3, requires the licensee to develop a performance-based leakage-testing program using the RG or other implementation document and referencing it in the plant TS. The submittal for TS revisions must also contain justification, including supporting analyses, if the licensee deviates from methods approved by the U.S. Nuclear Regulatory Commission (NRC or Commission) and endorsed in RG 1.163, "Performance-Based Containment Leak-Test Program." Option B specifies performance-based requirements and criteria for preoperational and subsequent leakage rate testing. These requirements are met by:

1. Type A tests to measure the containment system overall integrated leakage rate,

2. Type B pneumatic tests to detect and measure local leakage rates across pressure retaining leakage-limiting boundaries such as penetrations, and

3. Type C pneumatic tests to measure CIV leakage rates.

After the containment system has been completed and is ready for operation, Type A tests are conducted at periodic intervals based on the historical performance of the overall containment system to measure the overall integrated leakage rate. The leakage rate test results must not exceed the maximum allowable leakage (La) at design-basis loss-of-coolant accident (DBLOCA) pressure (Pa) with margin, as specified in the TSs. Option B also requires that a general visual inspection for structural deterioration of the accessible interior and exterior surfaces of the containment system, which may affect the containment leaktight integrity, be conducted prior to each Type A test and at a periodic interval between tests based on the performance of the containment system. Type B and Type C tests are performed based on the safety significance and historical performance of each boundary and isolation valve to ensure integrity of the overall containment system as a barrier to fission product release. The regulations in 10 CFR 50.36(c)(5), "Technical specifications," require, in part, the inclusion of administrative controls in TSs that are necessary to ensure operation of the facility in a safe manner. This LAR requests a change to a TS under the "Administrative Controls" section of the Fermi TSs.

Section 50.55a ,"Codes and standards," of 10 CFR contains the containment inservice inspection (ISi) requirements, which, in conjunction with the requirements of 10 CFR Part 50, Appendix J, ensure the continued leaktight and structural integrity of the containment during its service life. Section 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," paragraph (a)(1 ), states, in part, that the licensee:

        ... shall monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components, ... are capable of fulfilling their intended functions. These goals shall be established commensurate with safety and where practical, take into account industrywide operating experience.

NEI 94-01, Revision O (ADAMS Accession No. ML11327A025), provides methods for complying with the provisions of 10 CFR Part 50, Appendix J, Option B, and includes provisions that address the extension of the performance-based Type A test interval for up to 10 years, based upon two consecutive successful tests. Final Safety Evaluation (SE) for NEI Topical Report (TR) 94-01, Revision 2, "Industry Guideline For Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," and Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2, August 2007, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," dated June 25, 2008 (ADAMS Accession No. ML081140105), states that NEI 94-01, Revision 2 describes an acceptable approach for implementing the optional performance-based requirements of 10 CFR Part 50, Appendix J, Option B. The NRC staff concluded that NEI 94-01, Revision 2, is acceptable for referencing by licensees proposing to amend their containment leakage rate testing TSs, subject to the specific limitations and conditions listed in Section 4.1 of the SE. NEI 94-01, Revision 2-A (ADAMS Accession ML100620847), incorporates the regulatory positions stated in RG 1.163, and includes provisions for extending Type A test intervals up to 15 years. EPRI Report No. 1009325, Revision 2 1 , provides a generic assessment of the risks associated with a permanent extension of the integrated leak rate test {ILRT) surveillance interval to 15 years, and a risk-informed methodology/template to be used to confirm the risk impact of the ILRT extension on a plant-specific basis. Probabilistic risk assessment (PRA) methods are used, in combination with ILRT performance data and other considerations, to justify the extension of the ILRT surveillance interval. This is consistent with guidance provided in RGs 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," and 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," to support changes to surveillance test intervals. NEI 94-01, Revision 3-A, July 2012, provides guidance for extending Type C local leak rate test (LLRT) intervals beyond 60 months. TS 5.5.12.d.1 requires that Types A, B, and C test results must not exceed the La with margin. Option B also requires that a general visual inspection of the accessible interior and exterior surfaces of the containment system for structural deterioration, which may affect the 1 EPRI Report 1018243 is also identified as EPRI Report 1009325, Revision 2-A. This report is publicly available and can be found at www.epri.com by typing "1018243" in the search box.

containment leaktight integrity, be conducted prior to each Type A test and at a periodic interval between tests, based on the performance of the containment system. A Type A ILRT is currently required to be performed once every 1O years. The LAR asks to extend the interval to 15 years. The last ILRT was completed during Refueling Outage (RF)12 in November 1992, and using the proposed interval of 15 years, the next ILRT would have to be completed before the end of November 2022. 3.0 Technical Evaluation 3.1 Description of the Primary Containment The following excerpts from the March 22, 2016, LAR, Enclosure 1, Section 2.2, "Description of Fermi 2," provide a containment description: Fermi 2 utilizes a General Electric (GE) Boiling Water Reactor (BWR) Mark I primary containment structure. The containment consists of an inverted light-bulb-shaped steel liner vessel (drywell) and a torus-shaped suppression chamber. The primary containment consists of a drywell that houses the reactor pressure vessel (RPV), vent system connecting the drywell and the suppression chamber, reactor coolant recirculation system, vacuum relief system, isolation valves, and other primary components. The drywell is a steel pressure vessel that is enclosed in reinforced concrete for shielding purposes. The drywell is separated from the reinforced concrete structure by a gap of approximately 2 inches. This gap is filled with a compressible polyurethane material to allow for movement between the drywell and concrete. The polyurethane sheets are coated on both sides with an epoxy resin binder to prevent water leakage into the material. The bottom portion of the drywell shell is totally embedded in concrete. There is compacted sand at the lower portion of the transition zone between the upper freestanding section and the bottom embedded section of the drywell to allow for thermal expansion and to aid in the drainage of condensate that may accumulate in the 2-inch gap outside the Drywell. Four drain lines are used to remove any moisture in the sand cushion. The drywell vessel is provided with a removable head to facilitate refueling. Two bolted equipment hatches are provided for access into the drywell and one control rod drive (CAD) removal hatch is used for CRD replacement. There is also one double-door personnel airlock. The locking mechanism on each airlock door is designed to maintain a tight seal when the doors are subject to either external or internal pressure. The doors are mechanically interlocked so that neither door can be operated unless the other door is closed and locked. The drywell head and equipment hatches are designed with double gaskets with intermediate leak taps provided for leak testing. Provisions have been made to leak test the personnel air lock and door seals, equipment hatches, and CRD hatch. The Pressure Suppression Chamber, or torus, is a steel pressure vessel, in the shape of a torus, situated below and encircling the Drywell. The suppression chamber shell thickness is typically 0.587 inches above the horizontal centerline and 0.658 inches below the horizontal centerline, except at penetration locations

where it is locally thicker. The Suppression Chamber is supported vertically by inside and outside columns and by a saddle support that spans the inside and outside columns. The support system transmits dead weight and seismic and hydrodynamic loading to the reinforced-concrete foundation slab of the reactor building. Space is provided outside the chamber for inspection .... There are two 4-feet diameter manhole entrances with double-gasket bolted covers connected to the chamber by 4-feet diameter steel pipes. They are designed with leak testing connections. These access ports are closed when Primary Containment is required and opened when pressure suppression capability is no longer required. Eight circular vent pipes, 6 feet in diameter, form the connection between the drywell and the suppression chamber. These vent pipes are connected to a torus-shaped ring header, 4 feet, 3 inches in diameter, placed within the air space of the suppression chamber. The vent pipes are provided with two-ply expansion bellows to accommodate differential motion between the drywell and suppression chamber. These bellows have test connections that allow for leak testing and for determining that the passages between the two-ply bellows are not obstructed. Projecting downward from the torus-shaped ring header are 80 downcomer pipes, 24 inches in diameter, projecting from the ring header and terminating below the water surface in the suppression chamber. Primary containment penetrations carry piping, mechanical systems, and electrical wiring through the biological shield and primary containment vessel. These penetrations are classified as piping penetrations (sleeved and unsleeved), electrical service penetrations, mechanical system penetrations (traversing in-core probe penetrations), and access openings. The containment penetrations have the following design characteristics to maintain design containment integrity: Capability to withstand peak transient pressures, capability to withstand without failure the forces caused by impingement of the fluid from the rupture of the largest local pipe or connection, and the capability to accommodate without failure the thermal and mechanical stresses that may be encountered during all modes of operation. 3.2 Type A Integrated Leak Rate Test History The primary containment was designed for a maximum allowable La of 0.5 percent by weight of containment air weight per day at the calculated peak pressure, Pa. TS 5.5.12.b states that the "peak calculated containment internal pressure for the DBLOCA, Pa, is 56.5 psig [pounds per square inch guage]." Since 1989, a total of three ILRTs have been performed on the containment, all with "As Found" satisfactory results and these three ILRT test results were documented in the LAR. In addition, the licensee's August 11, 2016, response to an NRC RAI provided additional technical details about the ILRT test results. The test results are summarized in Table 1 below. The licensee completed its last Type A test on November 10, 2007. Prior testing confirmed that the containment structure leakage is acceptable with considerable margin with respect to the TS acceptance criterion of 1.0 La. The Appendix J acceptance criterion at 95 percent confidence level is 1.0 La= 0.50 weight percent/day at Pa. Since the last three Type A test results, as

shown in Table 1 below, were less than 0.5 weight percent/day, a test frequency of at least once per 1O years was justified in accordance with NEI 94-01, Revision 0. TABLE 1 DTE Electric Company Type A Test Historical Results Since 1989 Test Gauge Upper Correction for Total Acceptance Completion Pressure Confidence Type B & Type Leakage( 3l Criteria(6l, La Date during "As Limit Measured C Tests( 2l Found" ILRT Leakage( 1l (% weight/day) (% weight/day) (% weight/day) (psig) (% weight /day) 11/29/1989 n/a<4 > 0.285 0.033 0.318 0.375 11/01/1992 57.51<5> 0.212 0.032 0.244 0.375 11/10/2007 57.025<5 > 0.1168 0.0964 0.2132 0.375 Table 1 Notes: (1) BN-TOP-1 Upper Confidence Limit (UCL) - licensee RAI response dated August 11, 2016, SBPB RAl-3 (2) Sum of (minimum pathway leakage rate) MNPLR + N2 In-Leakage Correction + H20 Level Correction - licensee RAI response dated August 11, 2016, SBPB RAl-3 (3) Performance Leakage Rate (PLR) as defined by NEI TR 94-01, Revision 3-A, Sections 5.0 and 9.2.3 - licensee RAI response dated August 11, 2016, SBPB RAl-3 (4) Not provided in either the LAR or in response to the August 2016, RAI (5) Licensee RAI response dated August 11, 2016, SBPB RAl-3 and SBPB RAl-2 (6) Per TS 5.5.12.c NEI 94-01, Revision 3-A, Section 9.2.3, "Extended Test Intervals," stipulates that if previous Type A tests were performed at reduced pressure, then at least one of the two consecutive tests must be performed at a Pa. of 56.5 psig. Section 9.1.2, "Test Interval," states that the elapsed time between the first and the last tests in a series of consecutive passing tests used to determine performance must be at least 24 months apart. In the LAR, the licensee confirmed that the last two historical ILRTs were performed on November 1992, and November 2007, which met the 24-month criteria, and that both tests were performed at higher than the peak calculated design-basis internal accident pressure for the DBLOCA, Pa, of 56.5 psig. NEI 94-01, Revision 3-A, Section 9.2.3, does not stipulate that the licensee is to recalculate past Type A test results to demonstrate conformance with the definition of "performance leakage rate." The IRLT results from 1989, 1992, and 2007, demonstrate ample margin (i.e., ~ 36 percent) between each "as-found" ILRT value and La. Thus, the NRC staff did not request the licensee to reconstitute the 1989 Type A test results because the staff found the licensee's response acceptable. Based on the historical ILRT test results and the licensee's August 11, 2016, RAI response, the NRC staff concludes that the criteria in of NEI 94-01, Revision 3-A, Sections 9.1.2 and 9.2.3, are satisfied.

NEI 94-01, Revision 3-A, states the test standard is the American Nuclear Standards Institute/American Nuclear Society (ANSl/ANS)-56.8-2002. In an RAI dated July 14, 2016 (ADAMS Accession No. ML16197A005), the NRC staff requested a detailed breakdown of the results of the November 1, 1992, and November 10, 2007, tests to determine whether these results complied with the definition of "performance leakage rate (PLR)" as defined in NEI 94-01, Revision 3-A, Section 5.0. The licensee responded by providing a comprehensive summary of both the 1992 and 2007 Type A test results that demonstrated compliance with the NEI 94-01, Revision 3-A, definition of PLR. The details of the response are reflected in the notes to Table 1 above. The NRC staff finds the RAI response acceptable. TS 5.5.12.d.1 establishes the maximum limit for startup following completion of Type A testing at :::; 0. 75 La, which equals 0.375 percent of containment air weight per day. The primary containment was designed for a leakage rate La not to exceed 0.5 percent by weight of containment air per day at Pa. The licensee provided data that indicates there has been adequate margin to the "as-found" performance limit as required in TS 5.5.12.c. The NRC staff finds that the last two containment leakage rate results meet the TS 5.5.12.c design criterion of 0.5 percent La per day at the DBLOCA Pa. Since the last two Type A tests had "as-found" test results of less than 1.0 La, a test frequency of 15 years in accordance with NEI 94-01, Revision 3-A, and the conditions and limitations of NEI 94-01, Revision 2-A, is acceptable for Fermi 2. 3.3 lnservice Inspection Program The ILRT, LLRT, and ISi programs for the primary containment collectively ensure the leak-tightness and structural integrity of the containment. The containment ISi program implements the requirements of 10 CFR 50.55a. Fermi 2 follows the American Society of Mechanical Engineers (ASME), Section XI, Subsection IWE (not Subsection IWL), 2001 Edition with 2003 Addenda, as mandated and modified by 10 CFR 50.55a for the current 10-year interval. The containment concrete is used only to meet intended functions of shielding and structural support; it does not serve as a pressure retaining function. Therefore, the NRC staff concludes that Subsection IWL is not applicable to a BWR containment. Visual examinations are required per Surveillance Requirement 3.6.1.1.1, and leakage rate testing is required per TS 5.5.12. 3.4 ISi Program for Containment Metal Liner - IWE Fermi 2 follows ASME Section XI, Subsection IWE (requirements for Class MC and metal liners of Class CC components) of 2001 Edition through 2003 Addenda. A general visual inspection of 100 percent of accessible surface areas of the metallic containment and the moisture barriers are performed during the RF within inspection periods. Since a 15-year ILRT interval spans at least four IWE inspection periods, the performance of examinations in accordance with Subsection IWE assures at least three general visual examinations of containment pressure boundary will occur. The frequency of examinations performed in accordance with Subsection IWE satisfies the requirements of NEI 94-01, Revision 3-A, Section 9.2.3.2.

The Fermi 2 ISi schedule is shown in the following table: TABLE 2 IWE - 100 Percent Visual Inspections Interval Period Outage(s) 1st Interval 3rd RF12 3 , RF13 pt RF141, RF15 2 2nd Interval 2nd RF16 1 , RF17 3rd RF18 4 , RF19 pt RF20 4 , RF21 4 - ILRT 3rd Interval 2nd 5 RF22 , RF23 6 3rd RF24 5 , RF25 6 Table 2 Notes: 1 100 percent visual inspection of accessible interior/exterior primary containment shell completed and torus exterior and torus interior vapor region completed. 2 100 percent visual inspection of accessible interior torus wetted region completed. 3 100 percent visual inspection of accessible interior/exterior primary containment shell completed and torus interior/exterior wetted and vapor region completed. 4 100 percent visual inspection of accessible interior/exterior primary containment shell and torus interior/exterior wetted and vapor region required. 5 100 percent visual inspection of accessible interior/exterior primary containment shell, torus exterior, and torus interior vapor region required. 6 100 percent visual inspection of accessible interior torus wetted region required. Visual inspections of the containment during RF07 in the spring of 2000 identified a small (0.02 x 0.04 x 0.093 inch deep) pit at the interface of an I-beam with the containment steel liner. The corrosion was attributed to a screw and uncoated washer that were in contact with an uncoated portion of the drywell shell in a beam seat area. The screw and washer were removed, and the drywell in the area of the pit was coated in 2003. The drywell in the vicinity of the sand pocket is 1.5 inches thick. Mark I containment has three primary areas susceptible to degradation: (1) the moisture seal area where the concrete floor and containment steel liner meet, (2) the sand cushion area near the bottom of the drywell shell, and (3) the wet region inside the torus. During the first containment inspection interval during RF07, the moisture seal at the interface between the drywell concrete floor and the steel shell was removed. This was done to perform a detailed inspection of the liner in the seal area, repair areas of degradation in the seal, and as a preventive maintenance task. The inspection found no degradation to the drywell shell. The area was repainted and a new moisture seal was installed. In accordance with ASME,

Section XI, 2001 Edition with 2003 Addenda, the moisture barrier is inspected 100 percent each inspection period. During RF16 in spring of 2014, ultrasonic thickness measurements of the sand cushion region (approximately 50 percent) were performed, which indicated that there was no loss of material. The torus is inspected by certified personnel during alternate RFs. NRC Information Notice (IN) 2006-01, "Torus Cracking in a BWR Mark I Containment" (ADAMS Accession No. ML053060311 ), described a through-wall crack and its probable cause in the torus of a BWR Mark I containment. The cracking identified in the heat-affected zone at the high pressure core injection (HPCI) turbine exhaust pipe torus penetration was most likely initiated by cyclic loading due to condensation oscillation during HPCI operation. The Fermi 2 HPCI design has a turbine exhaust pipe sparger that prevents this condition from occurring. During RF15, broken blisters, mechanical damage, and pinpoint rust areas were identified and repaired in the wetted areas of the torus. In the vapor region, all flaking paint was removed from the torus ring header, torus vacuum breaker valves, nitrogen supply lines, monorail rail, and torus walkway and handrail. Flaking or cracked coating was removed, and a protective coating was re-applied to the torus shell. No pitting of the torus was identified during the 2012 RF15 inspections. The licensee stated that in the future, IWE examinations will be performed according to the examination schedule per ASME Section XI, Subsection IWE, Table IWE-2500-1. The licensee also stated that there are currently no primary containment surface areas that require augmented examinations in accordance with ASME Section XI, Subsection IWE-1240. 3.5 Type Band Type C Leak Rate Test History TS 5.5.12.d.1 requires that the leakage rate acceptance criterion be :s; 1.0 La. During the initial unit startup following testing, the leakage rate acceptance criteria is :s; 0.60 La for Type B and Type C tests and :s; 0. 75 La for Type A tests.

TABLE 3 Fermi 2 Type B and Type C Leak Rate Summation History Since 2003 RFO As Found Path Percentage of As Left Path Percentage of (scfh) 0.6 La (scfh) 0.6 La RF09 62.50 35.2% 74.83 42.1% (2003) RF10 84.80 47.7% 62.32 35.1% (2004) RF11 379.79 213.6% 67.2 37.8% (2006) (Note 1) RF12 359.83 202.4% 51.49 29.0% (2007) (Note 2) RF13 39.79 22.4% 75.48 42.5% (2009) RF14 46.71 26.3% 76.94 43.3% (2010) RF15 43.51 24.5% 80.38 45.2% (2012) RF16 44.54 25.1% 88.91 50.0% (2014) RF17 78.17 44.0% 74.46 41.9% (2015)

     =

scfh standard cubic feet per hour Note 1: LER [license event report] 2006-001, "Excessive Feedwater Check Valve Leakage at Containment Penetration" was issued. Note 2: LER 2007-001, "Excessive Feedwater Check Valve Leakage at Containment Penetration" was issued. In the NRC staff Table 3 shown above, Notes 1 and 2 identified two LERs that document the causes for the LLRT failures during RF11 in 2006 and RF12 in 2007, respectively. During RF11, LER 2006-001, "Excessive Feedwater Check Valve Leakage at Containment Penetration," dated May 24, 2006 (ADAMS Accession No. ML061520405), was issued to document that penetration X-9A minimum-pathway air leakage value was greater than the allowable containment leakage rate La. The failure was attributed to soft seat degradation on the feedwater line check valves, 821 OOF01 QA (inboard) and 821 OOF076A (outboard). For both the inboard and outboard check valves, the soft seats were replaced and the soft seat service time after RF11 was limited to two operating cycles. In addition, for the inboard check valve B2100F010A, the internal shaft was replaced and the alignment between the disc and the valve seat was adjusted. During RF12, LER 2007-001, "Excessive Feedwater Check Valve Leakage at Containment Penetration," dated December 6, 2007 (ADAMS Accession No. ML073460442), was issued to document a containment minimum pathway leak rate for a reactor feedwater line for penetration X-98 that exceeded the limiting conditions for operation in the plant's TSs. The check valve failures, 821 OOF01 OB (inboard) and 821 OOF076B (outboard), were primarily attributed to soft seat erosion by feedwater flow to the point that the seats were not providing an effective seal.

The corrective actions described in LER 2006-001 were taken for penetration X-9A, and this failure occurred before additional actions for other penetrations (i.e., penetration X-9B) were implemented. A preventive maintenance frequency to replace the soft seats every RF was established for the outboard feedwater check valves. In addition, the inboard isolation valves B21 OOF01 OA and B21 OOF01 OB were hard-seated during RF14 between October - December, 2010, and have been successfully "as-found" tested since. In the July 14, 2016, RAI, the NRC staff requested additional information about the corrective actions associated with the "cause of failure" for the valves listed against the penetration numbers identified in the LAR "Table 4.1-1: Type Band Type C LLRT Program Implementation Review of RF16 (2014)," and "Table 4.1-2: Type B and Type C LLRT Program Implementation Review of RF17 (2015)." The staff requested clarification on the causes of the repetitive failures, as well as whether there may be other failures of LLRTs associated with penetrations X-9A and X-9B since RF12 in 2007 that might suggest the corrective actions associated with LER 2007-001 may need further enhancement. The licensee responded to the staff in a letter dated August 11, 2016, letter (ADAMS Accession No. ML16225A493). The corrective actions for the two valves, C41 OO-F006, "Standby Liquid Control Outboard Check Valve," and C41 OO-F007, "Standby Liquid Control Inboard Check Valve," associated with penetration X-42 consisted of changing the frequency of the valves being disassembled, inspected, cleaned, and valve soft seats being replaced from every 9 years to every 6 years. Both of these valves are tested on a once per refuel outage frequency. The licensee, in the August 11, 2016, RAI response, provided an acceptable explanation for the repetitive failures associated with penetrations X-19 and X-42 in Table 5-1, Type C Air," and Table 5-2, Type C Water." The licensee's response indicated that the corrective actions of LER 2007-001 have been sufficient and that there have been no additional valve failures for penetration X-9A since 2007. However, there have been two additional valve failures since 2007 for penetration X-9B. Valve B21 OOF01 OB, "Feedwater Supply Line B Inboard Check Valve, failed its Type C test in 2009 during RF13, with a measured leakage of 55.20 scfh. Valve B21 OOF076B, "Feedwater Supply Line B Outboard Check Valve, failed its Type C test in 2014 during RF16, with a measured leakage of 11.66 scfh. Both of these valves have Type C test administrative leakage limits of 10.0 scfh. The licensee concluded that the LER 2007-001 requirement to replace the soft seat for the outboard feedwater check valves B21 OOF076A/B every refueling outage has been successful with only the one marginal failure. The inboard feedwater check valves, 821 OOF01 OA/B, had their soft seats removed and were hard-seated during 201 O in RF14. All subsequent "as-found" tests have been successful. As stipulated by NEI 94-01, Revision 0, these penetrations are tested every refuel outage. In summary, the licensee provided a comprehensive response to NRC's July 14, 2016, RAls regarding the cause of each of the LLRT repetitive failures and explained the corrective actions that were taken to prevent them from occurring in the future. The NRC staff concludes from reviewing the LAR, Section 3.2.2, Type Band C Testing," and the licensee's August 11, 2016, RAI response to SBPB RAl-5, that the licensee satisfied Section 10.2.3, "Type C Test Interval" of NEI 94-01, Revision 0, which requires type C tests to be performed at least once every 30 months until adequate performance is established.

Based on the LAR Enclosure 1, Table 4: "Fermi 2 Type B and Type C Leak Rate Summation History Since 2003," the NRC staff concludes:

• Excluding the outlier results of RF11 and RF12, the "as-found" minimum pathway leak rate average shows an average of 32.1 percent of 0.6 La with a high of 47.7 percent of 0.6 La (i.e., 0.286 La).
• The "as-left" maximum pathway leak rate average for shows an average of 40.8 percent of 0.6 La with a high of 50.0 percent of 0.6 La (i.e. 0.300 La).

Based on this data, the NRC staff concludes that excluding the outlying LLRT failures of RF11 and RF12, the aggregate results of the "as-found minimum path" and "as-left minimum path" for all the Type Band Type C tests from 2003 through 2015 demonstrate a history of successful testing, since the aggregate test results were significantly less than the Type Band Type C test TS limit of :::; 0.60 La contained in TS 5.5.12.d.1. Type B tests include electrical penetrations, vacuum breaker-electrical penetrations, drywell head, equipment hatches, personnel airlock (interior seal, exterior seal, equalizing valve), reactor vessel stabilization manhole 0-ring seals, expansion bellows, traverse in-core probe flanges, residual heat removal (RHR) test line orifices, RHR spool flange, butterfly valve flange, torus water management supply spool flanges, two-ply torus bellows, relief valve flanges, and RHR blind flanges. The percent of the total number of Type B tested components that are on a 120-month extended performance-based test interval is 93.6 percent (102 of 109). Seven penetrations cannot be placed in the Option B extended test interval program. This group of seven includes the personnel airlock, drywell head, two equipment hatches, CRD hatch, and two torus hatches. Type C tests include primary CIVs. Currently, 151 of 171 (84.4 percent) eligible Type C components are on an extended test interval of 60 months. This does not include the 28 valves that cannot be placed in the Option B program per RG 1.163 (such as main steam isolation valves (MSIVs), purge and vent valves, and feedwater check valves). Therefore, the NRC staff concludes that the percentage of Type Band Type C components on extended frequencies represents acceptable performance and supports allowing an extended test interval of up to 75 months for Type C-tested CIVs, in accordance with the guidance of NEI 94-01, Revision 3-A. The LAR did not provide sufficient information about how Fermi 2 satisfies NEI 94-01, Revision 0, Section 11.3.1, "Performance Factors," and Section 11.3.2, "Programmatic Controls". Both of these NEI report sections pertain to determining and implementing extended test intervals for Type Band Type C components. In its August 11, 2016, RAI response, the licensee stated that all Type B and Type C test results performed during the previous operating cycle, as well as the results of Type B and Type C tests performed during an outage, were assessed and documented in a post-outage test report. This response is acceptable to the NRC staff. In response to NRC's RAI question SBPB RAl-4, the licensee stated in its August 11, 2016, RAI response, that each post-outage report:

        ... includes evaluations that take into account the past test performance, service, frequency, design, and safety impact for Type Band Type C tested components

that are evaluated for frequency extension. The evaluations include containment isolation valve failures, the failure cause and corrective actions, and valves that exceeded established repair guidelines. The NRG staff concludes that the Fermi 2 procedures will be enhanced to incorporate the amount of understatement in the minimum pathway Type Band Type C summation, as well as continue to assess and document the plant's containment penetrations and valve performance in post-outage evaluations. Per NEI 94-01, Revision 3-A, Section 11.3.2, "Programmatic Controls," for extending "test intervals of greater than 60 months for a Type B or a Type C tested component," the post-outage report will also include "an estimate of the amount of understatement in the minimum pathway Type B and C summation," and "the reasoning and determination of the acceptability of the extension," as stipulated in Section 11.3.2. The NRG staff notes that the licensee followed RG 1.163, Regulatory Position C.2, which provides for an the extension of test intervals for Type C components up to 60 months. In the March 22, 2016, LAR, the licensee stated that Type B and Type C containment leakage rate testing program requires tests to detect and/or measure leakage across pressure retaining and leakage limiting boundaries and CIVs. As discussed in NUREG-1493 and NEI 94-01, Revision 3-A, Type B and Type C tests can identify the majority of all containment leakage paths. The LAR adopts the guidance in NEI 94-01, Revision 3-A, in place of NEI 94-0, Revision 0. A NRG staff review of the Type B and Type C test results from spring 2003 through fall of 2015 shows a large amount of margin between the actual "as-found" and "as-left" outage summations and the TS leakage rate acceptance criteria (that is, less than 0.6 La), as shown in Table 3 above. In summary, the NRG staff concludes as follows:

• Any lack of information or ambiguities contained in the LAR were adequately explained by the licensee in its August 11, 2016, RAI response.
• The licensee meets the guidance of both RG 1.163 and NEI 94-01, Revision 0.
• The cumulative Type B and Type C test results were below the acceptance limit of TS 5.5.12.d.1.
• The licensee has a corrective action program that appropriately addresses under-performing valves.

Accordingly, the staff finds that the licensee is effectively implementing the Type Band Type C leakage rate test program as required by 10 CFR 50, Appendix J, Option B. 3.6 NEI 94-01. Revision 2-A, Conditions The Fermi 2 containment is subject to the requirements set forth in 10 CFR 50, Appendix J, Option B, which requires that test intervals for Types A, B, and C testing be determined by using a performance-based approach. The Appendix J testing program plan is based on RG 1.163, which endorses NEI 94-01, Revision 0. The licensee's March 22, 2016, and August 11, 2016,

letters proposed to revise the Appendix J testing program by implementing the guidance contained in NEI 94-01, Revision 3-A, and the conditions and limitations of NEI 94-01, Revision 2-A. By letter dated June 25, 2008, the NRC published an SE with limitations and conditions for NEI 94-01, Revision 2. In the SE, the NRC staff concluded that NEI 94-01, Revision 2, describes an acceptable approach for implementing the optional performance-based requirements of Appendix J, and is acceptable for referencing by licensees proposing to amend their containment leakage rate testing TSs, subject to the limitations and conditions noted in Section 4.0 of the SE. NEI incorporated the June 25, 2008, NRC final safety evaluation report (SER) in NEI 94-01, Revision 2, which was subsequently issued as Revision 2-A on November 19, 2008. NEI 94-01 Revision 2-A, SE Section 4.1, "Limitations and Conditions for NEI TR 94-01, Revision 2," establishes limitations and conditions pertaining to deterministic requirements, while Section 4.2, "Limitations and Conditions for EPRI Report No. 1009325, Revision 2." establishes limitations and conditions pertaining to the plant's PRA analysis SE Section 4.1 includes provisions for extending the ILRT Type A interval to a maximum of 15 years, subject to the six limitations and conditions provided in the SE. The SE states that NEI 94-01, Revision 2, incorporates the regulatory positions found in RG 1.163. The leakage rate testing requirements of 10 CFR 50, Appendix J, Option B, for Types A, B, and C tests, and the containment ISi requirements mandated by 10 CFR 50.55a, together, ensure the continued leaktight and structural integrity of the containment during its service life. Type B testing ensures that the leakage rate of individual containment penetration components is acceptable. Type C testing ensures that individuals are leaktight. In addition, aggregate Type B and Type C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths. In the March 22, 2016, LAR, the licensee proposed to invoke NEI 94-01, Revision 3-A, along with the conditions and limitations of NEI 94-01, Revision 2-A, as the reference documents for Fermi 2 TS 5.5.12. Therefore, the licensee is also applying to extend the frequencies of the Type C performance-based test intervals beyond 60 months. The NRC staff has found that NEI 94-01, Revision 2-A, is acceptable for referencing by licensees proposing to amend their TSs to permanently extend the ILRT surveillance interval to 15 years, provided the following applicable six limitations and conditions are satisfied: Limitation and Condition 1 Limitation and Condition 1 specifies that for calculating the Type A leakage rate, the licensee should use the definition in NEI 94-01, Revision 2, in lieu of that in ANSl/ANS-56.8-2002. Licensee's Response to Limitation and Condition 1 Following NRC approval of this LAR, the licensee will use the definition in NEI 94-01, Revision 3-A, for calculating the Type A leakage rate, which is the same definition as in NEI 94-01, Revision 2.

Staff Assessment of Licensee's Response to Limitation and Condition 1 Section 3.2.9, "Type A test performance criterion, of ANSl/ANS-56.8-2002, defines the "performance leakage rate" and states: The performance criterion for a Type A test is met if the performance leakage rate is less than La. The performance leakage rate is equal to the sum of the measured Type A test UCL and the total as-left MNPLR of all Type B or Type C pathways isolated during performance of the Type A test. The June 25, 2008, NRC SE, Section 3.1.1.1 states: Section 5.0 of NEI TR 94-01 NEI TR 94-01, Revision 2, uses a definition of "performance leakage rate" for Type A tests that is different from that of ANSl/ANS-56.8-2002. The definition contained in NEI TR 94-01, Revision 2, is more inclusive because it considers excessive leakage in the performance determination. In defining the minimum pathway leakage rate, NEI TR 94-01, Revision 2, includes the leakage rate for all Type B and Type C pathways that were in service, isolated, or not lined up in their test position prior to the performance of the Type A test. Additionally, the NEI TR 94-01, Revision 2, definition of performance leakage rate requires consideration of the leakage pathways that were isolated during performance of the test because of excessive leakage in the performance determination. The NRC staff finds this modification of the definition of "performance leakage rate" used for Type A tests to be acceptable. NEI 94-01, Revision 2-A, Section 5.0, "Definitions, states: The performance leakage rate is calculated as the sum of the Type A upper confidence limit (UCL) and as-left minimum pathway leakage rate (MNPLR) leakage rate for all Type B and Type C pathways that were in service, isolated, or not lined up in their test position (i.e., drained and vented to containment atmosphere) prior to performing the Type A test. In addition, leakage pathways that were isolated during performance of the test because of excessive leakage must be factored into the performance determination. The performance criterion for Type A tests is a performance leak rate of less than 1.0 La. The NRC staff reviewed the definitions of "performance leakage rate" contained NEI 94-01, Revision 2, Revision 2-A, and Revision 3-A. The staff concludes that the definitions contained in all three revisions are identical. Therefore, the NRC staff concludes that because the definition found in Section 5.0 of NEI 94-01, Revision 2, for calculating the Type A leakage rate in the Fermi 2 "Primary Containment Leakage Rate Testing Program" is identical to that in NEI 94-01 Revision 3-A, the licensee adequately addressed "Limitation and Condition 1." Limitation and Condition 2 Limitation and Condition 2 stipulates that the licensee submit a schedule of containment inspections to be performed prior to and between Type A tests.

Licensee's Response to Limitation and Condition 2 In its March 22, 2016, LAA, in table captioned Table 2: Limitations/Conditions from NEI 94-01, Revision 2-A," the licensee stated: The frequency for performing general visual examinations of accessible interior and exterior surfaces of the containment for structural deterioration that may affect leak-tight integrity are performed as shown below in Table 5: ASME Class MC Examination Schedule for Fermi 2 (ASME Section XI, 2001 Edition with 2003 Addenda). The Fermi 2 IWE program is in its second 1O year interval. Staff Assessment of Licensee's Response to Limitation and Condition 2 NEI 94-01 Revision 2-A, SE Section 3.1.1.3, "Adequacy of Pre-Test Inspections (Visual Examinations), in part, states: NEI TR 94-01, Revision 2, Section 9.2.3.2, states that: "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." NEI TR 94-01, Revision 2, recommends that these inspections be performed in conjunction or coordinated with the examinations required by ASME Code, Section XI, Subsections IWE and IWL. The NRC staff finds that these visual examination provisions, which are consistent with the provisions of regulatory position C.3 of RG 1.163, are acceptable considering the longer 15 year interval. Regulatory Position C.3 of RG 1.163 recommends that such examination be performed at least two more times in the period of 1O years. The NRC staff agrees that as the Type A test interval is changed to 15 years, the schedule of visual inspections should also be revised. Section 9.2.3.2 in NEI TR 94-01, Revision 2, addresses the supplemental inspection requirements that are acceptable to the NRC staff. NEI 94-01, Revision 3-A, Section 9.2.1, "Pretest Inspection and Test Methodology," states: Prior to initiating a Type A test, a visual examination shall be conducted of accessible interior and exterior surfaces of the containment system for structural problems that may affect either the containment structure leakage integrity or the performance of the Type A test. This inspection should be a general visual inspection of accessible interior and exterior surfaces of the primary containment and components. 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. NEI 94-01, Revision 3-A, Section 9.2.3.2, "Supplemental Inspection Requirements," states: 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 NRC staff reviewed LAR, Enclosure 1, Section 3.3.1, "Containment lnservice Inspection Program (IWE)." The staff notes that LAR "Table 5: American Society of Mechanical Engineers (ASME) Class MC Examination Schedule for Fermi 2." indicates that for "Containment Surfaces" Item Number E1 .11, "Accessible Surface Areas," one 100 percent IWE examination is performed for each period in a 10-year interval. There are three periods per 10-year interval. Fermi 2 is currently in the second period of the second containment inspection interval. In the March 22, 2016, LAR, the licensee provided the following table: TABLE 4 Schedule of 100% IWE Visual Inspections Interval Period Outage(s) 1st Interval 3rd RF12 3 , RF13 2nd Interval pt RF141, RF15 2 2nd RF16 1, RF17 3rd RF18 4 , RF19 3rd Interval 1st RF20 4 , RF21 4 - ILRT 2nd RF22 5 , RF23 6 3rd RF24 5 , RF25 6 Table 4 Notes:

1. 100 percent visual inspection of accessible interior/exterior primary containment shell completed and torus exterior and torus interior vapor region completed.

2. 100 percent visual inspection of accessible interior torus wetted region completed.

3. 100 percent percent visual inspection of accessible interior/exterior primary containment shell completed and torus interior/exterior wetted and vapor region completed.

4. 100 percent visual inspection of accessible interior/exterior primary containment shell and torus interior/exterior wetted and vapor region required.

5. 100 percent visual inspection of accessible interior/exterior primary containment shell, torus exterior, and torus interior vapor region required.

6. 100 percent visual inspection of accessible interior torus wetted region required.

Fermi 2 completed its last Type A test during RF12 in November 2007. The next Type A test will be required to be completed before November 7, 2022, if the extension is approved. The licensee has scheduled this next Type A test to be performed during RF21 in September 2021, during period one of inspection interval three. According to Table 5 and Table 6 provided in the March 22, 2016, LAR, three 100-percent visual examinations of accessible areas of the containment are projected to be completed by the end of RF19 (i.e., the third period of the second inspection interval). Table 4 indicates for period 1 of inspection interval 3, the IWE 100-percent visual inspection of accessible interior/exterior primary containment shell and torus interior/exterior wetted and vapor region is projected for completion in conjunction with the ILRT of RF21. Based on these two projections, the NRC staff concludes that the licensee's submitted schedule to visually inspect accessible areas of the containment complies with the guidance contained in NEI 94-01, Revision 3-A, Sections 9.2.1 and 9.2.3.2, and, therefore, the provisions contained in NEI 94-01, Revision 2-A, Section 3.1.1.3, are satisfied. The NRC staff finds that the licensee adequately addressed Limitation and Condition 2. Limitation and Condition 3 Limitation and Condition 3 stipulates that the licensee address the areas of the containment structure potentially subjected to degradation. Licensee's Response to Limitation and Condition 3 In the March 22, 2016, LAR, "Table 2: Limitations/Conditions from NEI 94-01, Revision 2-A," the licensee stated: General visual examinations (general visual, VT-3, and VT-1 (if required)) of accessible interior and exterior surfaces of the class MC components, parts and appurtenances of the containment as well as a visual examination of the moisture barrier at the concrete-to-steel interface are performed. These include inspections of the downcomers, ring girders, torus shell (interior/exterior), vent lines, penetration sleeves, bolting, coatings, and penetration bellows (which are leak rate tested). In 2013, all four sand cushion drains lines from the sand cushion region were internally inspected with a boroscope to ensure free water flow. Periodic inspections of the drain lines indicate that there is no water present and a drain path through the lines has been verified. There are currently no primary containment surface areas that require augmented examinations in accordance with ASME Section XI, Subsection IW E-1240. Staff Assessment of Licensee's Response to Limitation and Condition 3 The NRC staff reviewed the following information provided in the March 22, 2016, LAR: (1) Section 3.3, "Containment Inspections";

(2) Section 3.3.1, "Containment lnservice Inspection Program (IWE)"; and (3) Section 3.4, "NRG Information Notice 92-20, "Inadequate Local Leak Rate Testing." NEI 94-01, Revision 2-A, SE Section 3.1.3, "Type A Test (ILRT), Type Band Type C Tests (LLRTs), and Containment In-Service Inspections (ISls)," states: In approving for Type A tests the one-time extension from 10 years to 15 years, the NRG staff has identified areas that need to be specifically addressed during the IWE and IWL inspections including a number of containment pressure-retaining boundary components (e.g., seals and gaskets of mechanical and electrical penetrations, bolting, penetration bellows) and a number of the accessible and inaccessible areas of the containment structures (e.g., moisture barriers, steel shells, and liners backed by concrete, inaccessible areas of ice condenser containments that are potentially subject to corrosion) ... Fermi 2 performs general visual examinations of the accessible surfaces of primary containment to assess the general condition of the primary containment surfaces. The Fermi 2 "Primary Containment lnservice Inspection Program," is based on ASME Code Section XI, Subsection IWE, and applies to the containment vessel. In the March 22, 2016, LAR, Section 3.3.1 "Containment lnservice Inspection Program (IWE)" states, in part: The scope of the Fermi 2 IWE program includes the free-standing steel containment vessel and its integral attachments, containment hatches, airlocks, moisture barrier, and pressure-retaining bolting. The components subject to examination are described in the Fermi 2 ISi-NOE Program. These include accessible surface areas, bolted connections, accessible surface areas of wetted or submerged, the vent system accessible areas, moisture barriers, and coatings. In Section 3.3 of the March 22, 2016, LAR, the licensee indicated that it follows the ASME, Section XI, Subsection IWE, 2001 Edition with 2003 Addenda, and the moisture seal at the interface between the drywell concrete floor and the steel shell is inspected 100 percent each ASME inspection period. Three period inspections are performed during each 10-year interval. In 2007 (i.e., RF07), the moisture seal was removed to facilitate inspection and as a preventative measure. This inspection found no degradation to the drywell shell. The liner seal area was repainted and a new moisture seal was installed. , Section 3.3.1, of the March 22, 2016, letter, states, in part: For the leak tightness of seals, gaskets and bolted connections, DTE will continue to perform inspections approved by the NRG on these components as described in the containment inspection program. Seals and gaskets undergo alternative testing in accordance with 10 CFR 50, Appendix J, Type B. Bolted connections are tested per Appendix J and are subject to a general visual inspection once each containment inspection period. DTE also performs post-maintenance Appendix J testing following any repair or disassembly of a component with a seal, gasket, or bolted connection.

The containment surfaces that may require augmented examination are performed in accordance with the licensee's "Primary Containment lnservice Inspection Program." The details of the examinations are provided in the licensee's March 22, 2016, LAR, "Table 5: ASME Class MC Examinations Schedule For Fermi 2," under the heading "Exam Cat." E-C and "Containment Surfaces Requiring Augmented Examination." There are currently no primary containment surface areas that require augmented examinations in accordance with ASME Section XI, Subsection IWE-1240. With respect to NRG IN 1992-20, "Inadequate Local Leak Rate Testing," pertaining to inadequate LLRTs of containment penetration bellows, the licensee determined that the bellows installed at Fermi 2 have a wire mesh between the plies that ensures an air gap for the adequate performance of Appendix J, Type B, testing. The NRG staff concludes that based on the information contained in the March 22, 2016, LAR, , Sections 3.3, 3.3.1, and 3.4, the licensee adequately addressed NEI 94-01, Revision 2-A, SE Section 3.1.3, and Limitation and Condition 3. The licensee also established its intent to satisfy the requirements of Limitation and Condition 3. Limitation and Condition 4 Limitation and Condition 4 specifies that the licensee address any tests and inspections performed following major modifications to the containment structure, as applicable. Licensee's Response to Limitation and Condition 4 In its table captioned "Table 2: Limitation and Condition from NEI TR 94-01, Revision 2-A," in Section 3.1, of Enclosure 1 to the March 22, 2016, letter, the licensee stated: There are currently no planned or anticipated major modifications to the Fermi 2 containment structure. The station design change process would address testing requirements for any future containment structure modifications. Staff Assessment of Licensee's Response to Limitation and Condition 4 NRG staff SE, Section 3.1.4, states: NEI TR 94-01, Revision 2, Section 9.2.4 of, states that: "Repairs and modifications that affect the containment leakage integrity require LLRT or short duration structural tests as appropriate to provide assurance of containment integrity following the modification or repair. This testing shall be performed prior to returning the containment to operation." Article IWE-5000 of the ASME Code, Section XI, Subsection IWE (up to the 2001 Edition and the 2003 Addenda), would require a Type A test after major repair or modifications to the containment. In general, the NRG staff considers the cutting of a large hole in the containment for replacement of steam generators or reactor vessel heads, replacement of large penetrations, as major repair or modifications to the containment structure. The Fermi 2 containment has been in service for approximately 35 years. By letter dated July 14, 2016, the NRG staff requested in RAI SBPB-1 that the licensee provide a history of repairs and modifications made to the containment structure to supplement the licensee's

response to Condition 4. In its August 11, 2016, letter, the licensee responded that there have been no repairs or modifications to the containment structure. Accordingly, the NRC staff concludes that no major modifications have taken place during the history of the containment and that there are no plans for anticipated major modifications to the containment structure. Any future containment structure major modifications will be adequately controlled and tested, consistent with the existing station design change process. Therefore, the NRC staff concludes that the licensee adequately addressed Limitation and Condition 4 by providing information that demonstrates it is inapplicable. Limitation and Condition 5 Limitation and Condition 5 specifies that 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 94-01, Revision 2, related to extending the ILRT interval beyond 15 years, the licensee must demonstrate to the NRC staff that it is an unforeseen emergent condition. Licensee's Response to Limitation and Condition 5 In its March 22, 2016, letter, "Table 2: Limitations/Conditions from NEI TR 94-01, Revision 2-A," the licensee stated: Fermi 2 acknowledges [Limitation and Condition 5] and accepts this NRC staff position, as communicated. Staff Assessment of Licensee's Response to Limitation and Condition 5 The June 25, 2008, NRC, SE, Section 3.1.1.2, states: 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 NEI 94-01, Revision 2, the licensee will have to demonstrate to the NRC staff that an unforeseen emergent condition exists. In the March 22, 2016, LAR, the licensee confirmed its understanding that any extension of the Type A test interval beyond the upper-bound performance-based limit of 15 years should be infrequent and that any requested permission (i.e., for such an extension) will demonstrate to the NRC staff that an unforeseen emergent condition exists. Based on the above, the NRC staff finds that the licensee adequately addressed Limitation and Condition 5.

Limitation and Condition 6 Limitation and Condition 6 specifies that for plants licensed under 10 CFR Part 52, applications requesting a permanent extension of the ILRT surveillance interval to 15 years should be deferred until after the construction and testing of containments for that design have been completed and applicants have confirmed the applicability of NEI 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, including the use of past containment ILRT data. Licensee's Response to Limitation and Condition 6 In its March 22, 2016, letter, "Table 2: Limitations/Conditions from NEI 94-01, Revision 2-A," in Section 3.1, the licensee stated that for condition 6: Not applicable. Fermi 2 is not licensed under 10 CFR Part 52. NRC Staff Assessment of Licensee's Response to Limitation and Condition 6 The NRC staff concludes that Limitation and Condition 6 does not apply because Fermi 2 is licensed under 10 CFR 50. 3.7 Conclusion Related to the Six Limitations and Conditions Listed in NEI 94-01. Revision 2-A, Section 4.1. of the NRC SE The NRC staff evaluated each of the six limitations and conditions listed above and determined that the licensee adequately satisfied all of the limitations and conditions identified in NEI 94-01, Revision 2-A, Section 4.1, of the NRC SE. Therefore, the NRC staff finds it acceptable for Fermi 2 to adopt the "conditions and limitations" of NEI 94-01, Revision 2-A, SE, Section 4.1, as part of the implementation documents listed in TS 5.5.12. NEI 94-01, Revision 3-A, "Conditions" Fermi 2 containment is subject to the requirements set forth in 10 CFR 50, Appendix J, Option B, which allows that test intervals for Type A, Type B, and Type C testing to be determined by using a performance-based approach. Currently, TS 5.5.12 is implemented in accordance with RG 1.163. The licensee's March 22, 2016, LAR, and its supplement, propose to revise TS 5.5.12 by replacing RG 1.163 with NEI 94-01, Revision 3-A, along with the conditions and limitations of NEI 94-01, Revision 2-A, to govern the test frequencies and the grace periods for Type A , B and Type C tests. NRC published an SE with limitations and conditions for NEI 94-01, Revision 3, by letter dated June 8, 2012 (ADAMS Accession No. ML121030286). In the SE, the NRC concluded that NEI 94-01, Revision 3, describes an acceptable approach for implementing the optional performance-based requirements of Appendix J, and is acceptable for reference by licensees proposing to amend their containment leakage rate testing TSs, subject to two conditions. The SE was incorporated into Revision 3 and subsequently issued as NEI 94-01, Revision 3-A, on July 31, 2012 In the March 22, 2016, LAR, the licensee proposes to use NEI 94-01, Revision 3-A, as the implementation document for Type B and Type C LLRT program.

In the March 22, 2016, LAR, the licensee indicated that it will meet the limitations and conditions of NEI 94-01, Revision 3-A, Section 4.0, of the SE. Accordingly, Fermi 2 will be adopting, in part, the testing criteria ANSI/ANS 56.8-2002 as part of its licensing basis. As stated in NEI 94-01, Revision 3-A, Section 2.0, "Purpose and Scope," where technical guidance overlaps between NEI 94-01, Revision 3-A, and ANSI/ANS 56.8-2002, the guidance in NEI 94-01, Revision 3-A, takes precedence. The NRC staff has found that the use of NEI 94-01, Revision 3-A, is an acceptable reference for use in licensee TSs to extend the Option B to 10 CFR Part 50, Appendix J, Type B and Type C test intervals beyond 60 months, provided the following two conditions are satisfied: Topical Report Condition 1 The June 8, 2012, NEI 94-01, Revision 3, SE, Section 4.0, Condition 1, stipulates that: NEI TR 94-01, Revision 3, is requesting that the allowable extended interval for Type C LLRTs be increased to 75 months, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total). The staff is allowing the extended interval for Type C LLRTs be increased to 75 months with the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit. In addition, a corrective action plan shall be developed to restore the margin to an acceptable level. The staff is also allowing the non-routine emergent extension out to 84-months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI 94-01, Revision 3. At no time shall an extension be allowed for Type C valves that are restricted categorically (e.g. BWR MSIVs), and those valves with a history of leakage, or any valves held to either a less than maximum interval or to the base refueling cycle interval. Only non-routine emergent conditions allow an extension to 84 months. Condition 1 identifies three issues that are required to be addressed: (1) The allowance of an extended interval for Type C LLRTs of 75 months requires that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit; (2) A corrective action plan is to be developed to restore the margin to an acceptable level; and (3) Use of the allowed 9-month extension for eligible Type C valves is only allowed for non-routine emergent conditions, but not for valves categorically restricted and other excepted valves. Licensee's Response to Condition 1 The licensee's response for Condition 1 is provided in the March 22, 2016, LAR, Section 3.1, of , "Table 1: Limitations/Conditions from NEI TR 94-01, Revision 3-A," where the licensee stated: Following NRC approval of this license amendment request, DTE will include margin between the Type B and Type C leakage rate summation and its

regulatory limit in the post-outage report and develop and maintain corrective actions to restore margin to acceptable level, as necessary. In addition, DTE will only extend the Type C test interval to 84 months for non-routine emergent conditions, subject to the limitations in Topical Report Condition. NRG Staff Assessment of Licensee's Response to Condition 1 The NRG staff reviewed the requirements of NEI 94-01, Revision 3-A, against the licensee's response for Condition 1. Based on its review, the NRG staff concludes that the licensee acknowledges and stated its intent to comply with all the requirements of Condition 1. Topical Report Condition 2 NRG SE dated June 8, 2012, Section 4.0, Condition 2, stipulates that: 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. Most of the primary containment leakage experienced has been attributed to penetration leakage and penetrations are thought to be the most likely location of most containment leakage at any time. The containment leakage condition monitoring regime involves a portion of the penetrations being tested each refueling outage, nearly all LLRT's being performed during plant outages. 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 1 or 2 or even 3 refueling outages. Type C tests involve valves which, in the aggregate, will show increasing leakage potential due to normal wear and tear, some predictable and some not so predictable. Routine and appropriate maintenance may extend this increasing leakage potential. Allowing for longer intervals between LLRTs means that more leakage rate test results from farther back in time are summed with fewer just tested penetrations and that total used to assess the current containment leakage potential. This leads to the possibility that the LLRT totals calculated understate the actual leakage potential of the penetrations. 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 conservatively accounted for. 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 the trending specified in NEI 94-01, Revision 3, Section 12.1. When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B & C total, and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the

extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations. Condition 2 identifies two issues that are required to be addressed: (1) Extending the Type C 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 the trending specified in NEI 94-01, Revision 3, Section 12.1; and (2) When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the Primary Containment Leakage Rate Testing Program trending or monitoring must include an estimate of the amount of understatement in the Type B and Type C total, and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations. Licensee's Response to Condition 2 The "DTE Response" for Condition 2 is captured in the March 22, 2016, LAR, Section 3.1, "Table 1: Limitations/Conditions from NEI TR 94-01, Revision 3-A," Enclosure 1: Following NRC approval of this license amendment request, if routinely scheduling the LLRT valve interval beyond 60-months and up to 75-months, DTE will include in the post outage report an estimate of the amount of under-statement in the Type Band C total, and the reasoning and determination of the acceptability of the extension to demonstrate that the LLRT totals calculated represent the actual leakage potential of the penetrations. NRC Staff Assessment of Licensee's Response to Condition 2 The NRC staff reviewed the requirements of NEI 94-01, Revision 3-A, and concludes that the licensee acknowledges and stated its intent to comply with all the requirements of Condition 2, including the post-outage report trending requirements in Section 12.1 of NEI 94-01, Revision 3-A. NEI 94-01, Revision 3-A, SE Section 4.0, "Conclusion" Based on the evaluation of the two conditions in Section 4.0 of the NRC, SE, in NEI 94-01, Revision 3-A, the staff determined that the licensee adequately addressed both conditions. Therefore, the staff finds it acceptable for Fermi 2 to adopt NEI 94-01, Revision 3-A, as the implementation document in TS 5.5.12.a. The NRC staff reviewed the Type A, Type B, and Type C leakage test results related to the licensee's proposal to extend 10 CFR 50, Appendix J, test intervals. The ILRT results provided in the March 22, 2016, LAR, Section 3.2.1, "Type A Testing," indicate that the previous two consecutive Type A tests were successful, with containment performance leakage rates less than the maximum allowable rate of 0.5 percent containment air weight per day ( 1.0 La at Pa) and less than the Type A test TS limit of ::> 0. 75 La contained in TS 5.5.12.d.1.

Therefore, the NRC staff finds that the performance history of Type A tests supports extending the current ILRT inteNal on a permanent basis to 15 years, as permitted by NEI 94-01, Revision 3-A, and the conditions and limitations of NEI 94-01, Revision 2-A. The NRC staff also reviewed the local leak rate summaries contained in the March 22, 2016, LAR, Enclosure 1, "Table 4: Fermi 2 Type B and Type C Leak Rate Summation History Since 2003," and notes that the aggregate results of the "as-found min path" and "as-left min path" for all the recent (i.e., since RF13 in 2009) Type Band Type C tests are less than the Type Band Type C test TS limit of::;; 0.60 La contained in TS 5.5.12.d.1. The NRC staff reviewed the corrective actions identified in the March 22, 2016, LAR, Enclosure 1, Section 3.2.2, "Type B and C Testing," taken for the valves that failed the Type C LLRT program tests during RF11 in 2006 and RF12 in 2007, and concludes that adequate corrective action for the failed feedwater line check valves has been performed. Therefore, the NRC staff finds that the licensee is effectively implementing the Type Band Type C leakage rate test program, as required by 10 CFR 50, Appendix J, Option B. Accordingly, the staff finds that the performance history of Type B and Type C tests supports extending the current Type C test inteNal to 75 months as permitted by NEI 94-01, Revision 3-A. 3.8 Probabilistic Risk Assessment of the Proposed Extension of ILRT and LLRT Test lnteNals 3.8.1 Introduction 3.8.2 Probabilistic Risk Assessment Background Section 9.2.3.1, "General Requirements for ILRT lnteNal Extensions beyond Ten Years," of NEI 94-01, Revision 3-A (ADAMS Package Accession No. ML122210254), states that plant-specific confirmatory analyses are required when extending the Type A ILRT inteNal beyond 10 years. Section 9.2.3.4, "Plant-Specific Confirmatory Analyses," of NEI 94-01, Revision 3-A states that the assessment should be performed using the approach and methodology described in EPRI Report No. 1018243, "Risk Impact Assessment of Extended Integrated Leak Rate Testing lnteNals." 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 inteNal. In the SE dated June 25, 2008 (June 2008 SE), the NRC staff also found the methodology in EPRI Report No. 1009325, Revision 2, acceptable for referencing by licensees proposing to amend their TSs to permanently extend the ILRT inteNal to 15 years, provided certain conditions are satisfied. These conditions, set forth in Section 4.2 of the June 2008 SE stipulate that:

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

2. The licensee submits documentation indicating that the estimated risk increase associated with permanently extending the ILRT suNeillance inteNal to 15 years is small, and consistent with the clarification provided in Section 3.2.4.5 of this SE [June 2008]. 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 CCFP should be defined as a value marginally greater than that accepted in previous one-time 15-year ILRT extension requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage point. While acceptable for this application, the NRC staff is not endorsing these threshold values for other applications. Consistent with this limitation and condition, EPRI Report No. 1009325 will be revised in the "-A" version of the report, to change the population dose acceptance guidelines and the CCFP guidelines.

3. The methodology in EPRI Report No. 1009325, Revision 2, is acceptable except for the calculation of the increase in expected population dose (per year of reactor operation). In order to make the methodology acceptable, the average leak rate for the pre-existing containment large leak rate accident case (accident case 3b) used by the licensees shall be 100 La instead of 35 La.

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

3.8.3 Plant-Specific Risk Evaluation The licensee performed a risk impact assessment for extending the Type A containment ILRT interval to once in 15 years. The risk assessment was provided in Enclosure 4, "Evaluation of Risk Significance of Permanent ILRT Extension," in the March 22, 2016, LAR. In Section 1.0 "Purpose," of Enclosure 4 to the LAR, the licensee stated that the plant-specific risk assessment follows the guidance in: (1) NEI 94-01, Revision 2; the methodology described in EPRI TR 1009325, Revision 2-A; (2) NRC regulatory guidance outlined in RG 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (ADAMS Accession No. ML100910006); and (3) NRC regulatory guidance outlined in RG 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" (ADAMS Accession No. ML090410014). (4) Methodology from Calvert Cliffs Nuclear Power Plant to assess the risk from undetected containment leaks due to steel liner corrosion. In the June 2008 SE, Section 4.2 "Limitations and Conditions for EPRI Report No. 1009325, Revision 2," the NRC staff provides the four conditions that must be met in order permanently extend the ILRT surveillance interval to 15 years.

Condition 1 The licensee submits documentation indicating that the technical adequacy of their PRA is consistent with the requirements of RG 1.200 relevant to the ILRT extension application. NRC Staff Assessment of Condition 1 Consistent with the information in Regulatory Issue Summary 2007-06, NRC Regulatory Issue Summary 2007-06 Regulation Guide 1.200 Implementation" (ADAMS Accession No. ML070650428), the NRC staff used Revision 2 of RG 1.200 to assess technical adequacy of the PRA used to support risk-informed applications received after March 2010. Section 3.2.4.1 of the June 2008 SE states that Capability Category I of the ASME PRA standard shall be applied as the standard for assessing PRA quality for IRLT 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. In the March 22, 2016, LAR, "Appendix A - PRA Model Technical Adequacy," the licensee stated that the Fermi 2 V1 O model is the most recent evaluation of the risk profile at Fermi 2 for the ILRT application of internal events, including flooding at power operations. In Section A.1 "PRA Maintenance and Update," the licensee describes its process used for controlling the model and for ensuring that the model reflects the as-built, as-operated plant. The licensee also describes Fermi 2's process for continued PRA maintenance and update, including procedures for regularly scheduled and interim PRA model updates and for tracking issues identified as potentially affecting the PRA model. The NRC staff finds the Fermi PRA maintenance and update method acceptable. The licensee stated in the March 22, 2016, LAR, Appendix A, Section A.4, that the last full scope peer review of the internal events and internal flooding PRA was performed by the Boiling Water Reactor Owner's Group in August 2012, against the ASME PRA Standard (ASME/ANS RA-Sa-2009) and RG 1.200, Revision 2. The peer review performed on the Fermi 2 draft V9 PRA model was an evaluation of all supporting requirements of the ASME/ANS standards for all upgrades. The peer review team identified ten finding-level Facts and Observations (F&Os) which were all later determined to be PRA maintenance. The resolution of these peer review findings was incorporated into the final Fermi 2 V9 revision of the PRA . The licensee also stated that the most recent and current revision (Fermi 2 V10) of the PRA model was an update to the Human Reliability Analysis (HRA) dependency analysis. The licensee further stated that a focused scope peer review was performed on the HRA dependency analysis update in February 2014, which concluded that all reviewed supporting requirements met Capability Category II. The NRC staff reviewed the 1O finding level F&Os from the full scope peer review and their proposed resolution submitted by the licensee in Table A-2 of Enclosure 3 to the LAR, and concluded that Fermi 2 appropriately addressed the F&Os because there were no new methodologies or significant changes in scope or capability that affected the significant accident sequence as it pertains to the ILRT extension application. Section 3.2.4.2 of the June 2008 SE states: 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. The licensee, in the March 22, 2016, LAR, Appendix A, Section A.2, "Plant Changes not yet incorporated in to the PRA Model," stated that a review of the plant modifications and changes were performed following the plant refueling outage on November 2015, and concluded that currently there were no significant plant changes (design or operational practices) that have not yet been incorporated in the PRA models. The NRC staff concludes that the current PRA model is appropriate with respect to the current plant configuration and for the ILRT extension application. The NRC staff reviewed the licensee's assessed impact from external events provided in the March 22, 2016 LAR, Enclosure 4, Section 5.4.2 "Potential Impact from External Events". External hazards were evaluated against the "Individual Plant Examination of External Events {IPEEE) for Severe Accident Vulnerabilities - 10CFR 50.54(f) (Generic Letter (GL) 88-20, Supplement 4" submittal in response to the NRC IPEEE program (ADAMS Accession No. ML031150485). The assessment included the risk contribution from internal fire and seismic events. The licensee stated that the IPEEE evaluation of external hazards screened out 36 external hazards, including high winds, tornadoes, external floods, transportation accidents, nearby facility accidents, and release of onsite chemicals. The licensee further stated that this external events screening evaluation was peer reviewed in April 2014, using ASME/ANS RA-Sb-2009 and RG 1.200, Revision 2, and that all supporting requirements were met at Capability Category II. The NRC staff reviewed the Fermi 2 assessment and concludes that it meets the criteria provided in GL 88-20. Appendix D, "Seismic Core-Damage Frequencies," in "NRC Generic Issue 199 (Gl-199), Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants, Safety/Risk Assessment," August 201 O (ADAMS Package Accession No. ML100270582), published probabilistic seismic hazard for existing plants in the central United States. Since the IPEEE seismic margin analysis provided no quantification of seismic risk, Fermi 2 used a seismic CDF of 4.2E-06/year as provided in the Appendix D of Gl-199. NRC staff finds that the Fermi 2 use of 4.2E-06/year CDF is acceptable. The fire risk was calculated during IPEEE using "Fire-Induced Vulnerability Evaluation (FIVE), EPRI TR-100370 Final Report," May 1, 1992, which describes the fire-induced vulnerabilities evaluations and quantitative screening technique for fire analysis. Although the results of IPEEE studies included in the LAR provide insights on the risk from fire, the NRC staff considers results obtained from the IPEEE FIVE methodology to be outdated. Nevertheless, these IPEEE studies do provide an order of magnitude estimate for risk contribution from fire, as expected for this application. The NRC staff has found the use of the IPEEE study in assessing the impact from external events on the ILRT extension applications to be acceptable, once those studies are evaluated periodically, considering new information. However, the LAR did not provide any information to indicate a reassessment of IPEEE results based on new information for fire risk.

The NRC staff finds that the use of the data from the IPEEE FIVE method has no impact on the conclusions of this SE because sufficient margin exists between the risk acceptance criteria and the risk associated with the ILRT extension. Therefore, the staff finds the licensee's analysis of the impact of external events acceptable for the ILRT application. Staff Conclusion on Condition 1 The licensee submitted appropriate documentation that indicated the technical adequacy of their PRA is consistent with the requirements of RG 1.200 relevant to the ILRT extension application. The licensee's internal events PRA meets the requirements of the current ASME PRA standard and Revision 2 of RG 1.200. The licensee also included a quantitative assessment of the contribution of external events, including the effects of internal fires and seismic events. Therefore, the NRC staff concludes that the internal events PRA model used by the licensee is of sufficient technical adequacy, and the insights obtained from IPEEE studies are sufficient, to support the evaluation of changes to ILRT frequencies. Accordingly, the first condition is met. Condition 2 The licensee submits documentation indicating that the estimated risk increase associated with permanently extending the ILRT surveillance interval to 15 years is small, and consistent with the clarification provided in Section 3.2.4.5 of this SE [June 2008]. 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 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 conditional containment failure probability (CCFP) be less than or equal to 1.5 percentage point. While acceptable for this application, the NRC staff is not endorsing these threshold values for other applications. Consistent with this limitation and condition, EPRI Report No. 1009325 will be revised in the "-A" version of the report, to change the population dose acceptance guidelines and the CCFP guidelines. NRC Staff Assessment of Condition 2 The licensee submitted 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 clarification provided in Section 3.2.4.5 of June 2008 SE; and EPRI Report No. 1009325, Revision 2. 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. Fermi 2 does not rely on containment over-pressure for ECCS performance. Thus, NRC staff concludes that the associated risk metrics are LERF, population dose, and CCFP. The licensee reported the results of the plant-specific risk assessment in Section 3.5.3 of to the March 22, 2016, LAR. Details of the risk assessment are provided in of the LAR. The reported risk impacts are based on a change in test frequency from three tests in 10 years (the test frequency under 10 CFR 50, Appendix J, Option A) to one test in 15 years. The following conclusions can be drawn from the licensee's analysis associated with extending the Type A ILRT frequency:

1. The reported increase in LERF for internal events is 1.27E-08/year. The increase in LERF for combined internal and external events is 1.90E-07/year. The risk contribution

from external events includes the effects of internal fires and seismic. This change in risk is considered to be "small" (i.e., between 1E-06/year and 1E-07/year) per the acceptance guidelines in RG 1.174. An assessment of total baseline LEAF is required to show that the total LEAF is less than 1E-05/year. The licensee estimated the total LEAF for internal and external events as 7.88E-06/year. The total LEAF, given the increase in ILRT interval, is below the acceptance guideline of 1E-05/year in RG 1.174 for a "small" change.

2. The reported change in Type A ILRT frequency from three in 1O years to once in 15 years results in a reported increase in the total population dose of 1.14E-4 person-rem/year. The reported increase in total population dose is defined in Section 3.2.4.6 of the June 2008 SE as "a small increase in population dose as 0.75 person-rem per year or less." The information provided by Fermi is below the values found in EPRI Report 1009325, Revision 2-A. 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 1O years to once in 15 years is 0.73 percent. This value is below the acceptance guidelines in Section 3.2.4.6 of the NRC SE for NEI 94-01, Revision 2.

NRC Staff Conclusion on Condition 2 Based on the risk assessment results, the NRC staff concludes that the increase in LEAF is small and consistent with the acceptance guidelines of RG 1.174, 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 LAR. 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 is met. Condition 3 The methodology in EPRI Report No. 1009325, Revision 2, is acceptable except for the calculation of the increase in expected population dose (per year of reactor operation). In order to make the methodology acceptable, the average leak rate for the pre-existing containment large leak rate accident case (accident case 3b) used by the licensees shall be 100 La instead of 35 La. NRC Staff Assessment of Conclusion on Condition 3 The third condition stipulates that in order to make the methodology in EPRI Report 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 3 of Enclosure 4 to the LAR, the licensee used 100 La as the average leak rate for the preexisting containment large leakage rate accident case (accident case 3b) in the plant-specific risk assessment. Accordingly, the NRC staff finds that the third condition is met.

Condition 4 An LAR is required in instances where containment over-pressure is relied upon for ECCS performance. NRC Staff Assessment of Conclusion on Condition 4 The fourth condition stipulates that in instances where containment over-pressure is relied upon for ECCS performance during design-basis accidents, a LAR is required to be submitted. Since the PRA model credits over-pressure in long-term loss of containment heat removal scenarios, the licensee performed a sensitivity study and calculated an increase in CDF of 5.25E-8/year and an increase in LERF of 3.54E-8/year. These changes in risk are considered to be "very small" for both CDF and LERF per the acceptance guidelines of RG 1.174. The NRC staff concludes that containment over-pressure is not required to support ECCS performance in the Fermi 2 design basis. Accordingly, the NRC staff concludes that the fourth condition is not applicable. 3.8.4 Probabilistic Risk Assessment of the Proposed Extension of ILRT and LLRT Test Intervals Risk Evaluation Conclusion The NRC staff reviewed the risk evaluation provided as Enclosure 4 to the March 22, 2016, LAR submittal and determined that the licensee has adequately addressed each of the four conditions of EPRI Report No. 1009325, Revision 2. Therefore, NRC staff concludes that there is sufficient safety margin to implement the TS changes as submitted. 3.9 Licensee's Proposed Changes The first change inserts "and" before "(b)" and removes from paragraph 2.D of the renewed facility operating license, the exemption from containment air lock testing requirements which states: (c) an exemption from the requirement of Paragraph lll.D.2(b)(ii) of Appendix J, the testing of containment air locks at times when containment integrity is not required (Section 6.2.7 of SSER [supplemental safety evaluation report] (SSER) #5). Item (c) in paragraph 2.D of RFOL currently grants an exemption from Option A requirements in Part 50, Appendix J. License Amendment No. 108, which was issued in August 1996, approved the licensee's adoption of Option B of Appendix J for this testing. Accordingly, the NRC staff finds the licensee's proposed deletion of Item (c) of paragraph 2.D of the Fermi 2 operating license and the insertion of "and" before "(b)" in the same paragraph, acceptable. The second change would revise TS 5.5.12 by replacing the reference to RG 1.163 with a reference to NEI 94-01, Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A, to implement the performance-based leakage testing program in accordance with 10 CFR 50, Appendix J, Option B.

The licensee proposes to remove TS 5.5.12.a:

         ... guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, with the exception of approved exemptions to 10 CFR 50, Appendix J, and as modified by the following exception to NEI 94-01, Rev[ision] 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J."

The licensee requests replacement with the following: Nuclear Energy Institute (NEI) 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 2012, and the limitations and conditions specified in NEI 94-01, Revision 2-A, dated October 2008. The revised TS 5.5.12.a would state as follows: A program shall be established to implement the leakage rate testing of the primary 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 Nuclear Energy Institute (NEI) 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 2012, and the limitations and conditions specified in NEI 94-01, Revision 2-A, dated October 2008. The NRC staff reviewed the requested changes, from the perspective of deterministic considerations with regard to containment leaktight integrity and determined that the licensee was consistent with the guidance contained in both NEI 94-01, Revision 2-A, and NEI 94-01, Revision 3-A. The licensee justified the proposed changes by demonstrating adequate performance of the containment based on: (a) the historical plant-specific containment leakage testing program results, (b) the containment ISi program results, and (c) a plant-specific risk assessment, therefore, the staff finds the second change acceptable. The third change would also delete the following language from TS 5.5.12.a: The first Type A test after the October 1992 test shall be performed no later than October 2007. The NRC staff finds that the proposed deletion of the text in TS 5.5.12.a is appropriate because the first Type A test was performed on November 1, 1992 and the extension period expired in 2007. Accordingly, the staff approves the deletion of the indicated text TS 5.5.12.a. 3.10 CONCLUSION In the March 22, 2016, LAR and its supplement dated August 11, 2016, the licensee proposed to extend the current performance-based Type A test interval to 15 years by revising its TS to incorporate NEI 94-01, Revision 3-A, and the conditions and limitations of NEI 94-01, Revision 2-A, as the implementation documents in TS 5.5.12. This change would allow Fermi 2 to conduct the next Type A test on or before November 7, 2022, instead of the current due date of November 9, 2017.

Consistent with the guidance in NEI 94 01, Revision 3-A, and the conditions and limitations of NEI 94-01, Revision 2-A, the licensee justified the proposed change by demonstrating adequate performance of its containment based on: (a) plant-specific containment leakage testing program results, (b) containment ISi results, and (c) a plant-specific risk assessment. Based on the NRC staff's review of the licensee's March 22, 2016, LAR, and the supplemental information provided in the RAI response letter dated August 11, 2016, and the regulatory and technical evaluations above, the staff finds that the licensee has addressed the applicable NRC conditions to demonstrate acceptability of adopting NEI 94-01, Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A, as the 10 CFR 50, Appendix J, Option B, implementation documents. The NRC staff reviewed the proposed changes to verify the revised program description continues to contain the appropriate administrative controls for the Containment Leak Rate Testing Program. The NRC staff concludes that the revised TS continue to provide the appropriate administrative controls to ensure that the requirements of 10 CFR 50.36(c)(5) continue to be satisfied. The NRC staff also finds that the licensee adequately implemented its primary containment leakage rate testing program (i.e., Types A, B, and C leakage tests) for the Fermi 2 containment. The results of past ILRTs and recent LLRTs demonstrate acceptable performance of the containment and demonstrate that the structural and leaktight integrity of the containment structure is being adequately maintained. The NRC staff also finds that the structural and leaktight integrity of the containment will continue to be monitored and maintained if Fermi 2 adopts NEI 94-01, Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A, as the 10 CFR 50, Appendix J, Option B, implementation documents. Accordingly, the NRC staff determined that there is reasonable assurance that the structural and leaktight integrity for the containment will continue to be maintained, without undue risk to public health and safety, if the current Type A test intervals are extended to 15 years, and if the current Type C test intervals are extended to 75-months. The NRC staff concludes that it is acceptable for Fermi 2 to: (1) Revise TS 5.5.12 to include NEI 94-01, Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A, as the 10 CFR 50, Appendix J, Option B, implementation documents; (2) Delete the expired Type A test extension in TS 5.5.12.a; (3) Extend on a permanent basis the Type A test interval up to 15 years; (4) Extend the Type C test intervals up to 75-months; and (5) Delete exemption in License Condition 2.D(c) in the Fermi 2 Operating License

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Michigan State official was notified on August 17, 2016 of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes the surveillance requirements. The 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 (81 FR 36613, dated June 6, 2016). The amendment also makes corrective or other minor changes to the license. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (c)(1 O)(v). 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: David A. Nold Dan Hoang Jerrod Demers Date of issuance:March 9, 2O1 7

ML16351A460 *by internal memos dated OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DRA/APLA/BC* NRR/DE/EMCB/BC* NRR/DSS/SBPB/BC* NAME SGoetz SRhorer SRosenberg Yli RDennig (LRonewicz for) DATE 12/28/2016 12/22/2016 10/21/2016 9/12/2016 10/04/2016 OFFICE NRR/DSS/STSB/BC OGG NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME AKlein MYoung DWrona SGoetz (KGreen for) DATE 1/18/2017 3/8/2017 3/9/2017 3/9/2017}}