ML17251A211

Issuance of Amendment No. 307 Revision of Steam Generator Technical Specifications to Reflect Adoption of Technical Specifications Task Force Traveler TSTF-510 (CAC MF9654; EPID L-2017-LLA-0222)

ML17251A211
Person / Time
Site:	Arkansas Nuclear
Issue date:	10/10/2017
From:	Thomas Wengert Plant Licensing Branch IV
To:	Entergy Operations
Wengert T, NRR/DORL/LPLIV, 415-4037
References
CAC MF9654, EPID L-2017-LLA-0222
Download: ML17251A211 (16)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 10, 2017 ANO Site Vice President Arkansas Nuclear One Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802

SUBJECT:

ARKANSAS NUCLEAR ONE, UNIT 2 - ISSUANCE OF AMENDMENT RE:

REVISION OF STEAM GENERATOR TECHNICAL SPECIFICATIONS TO REFLECT ADOPTION OF TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER TSTF-510 (CAC MF9654; EPID L-2017-LLA-0222)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 307 to Renewed Facility Operating License No. NPF-6 for Arkansas Nuclear One, Unit 2 (AN0-2). The amendment consists of changes to the technical specifications (TSs) in response to your application dated April 24, 2017.

The amendment incorporates the guidance of Technical Specifications Task Force (TSTF),

Change Traveler TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection." The guidance of TSTF-510 revises TS 3.4.17, "Steam Generator (SG) Tube Integrity," TS 5.5.9, "Steam Generator (SG) Program," and TS 5.6.7, "Steam Generator Tube Inspection Report," of the Improved Standard Technical Specifications that are applicable to AN0-2.

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

Sincerely,

~'~~

Thomas J. Wen~Jt, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-368

Enclosures:

1. Amendment No. 307 to NPF-6

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY OPERATIONS, INC.

DOCKET NO. 50-368 ARKANSAS NUCLEAR ONE, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 307 Renewed License No. NPF-6

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

A. The application for amendment by Entergy Operations, Inc. (the licensee), dated April 24, 2017, 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 license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-6 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 307, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. The license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance.

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

Attachment:

Changes to the Renewed Facility Operating License No. NPF-6 and Technical Specifications Date of Issuance: October 1 O, 2O1 7

ATTACHMENT TO LICENSE AMENDMENT NO. 307 RENEWED FACILITY OPERATING LICENSE NO. NPF-6 ARKANSAS NUCLEAR ONE, UNIT 2 DOCKET NO. 50-368 Replace the following pages of the Renewed Facility Operating License No. NPF-6 and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Operating License REMOVE INSERT 3 3 Technical Specifications REMOVE INSERT 3/4 4-6 3/4 4-6 6-8 6-8 6-9 6-9 6-14 6-14 6-22 6-22

3 (4) EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) EOI, pursuant to the Act and 10 CFR Parts 30 and 70 to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

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

(1) Maximum Power Level EOI is authorized to operate the facility at steady state reactor core power levels not in excess of 3026 megawatts thermal. Prior to attaining this power level EOI shall comply with the conditions in Paragraph 2.C.(3).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 307, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

Exemptive 2nd paragraph of 2.C.2 deleted per Amendment 20, 3/3/81.

(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following issuance of the renewed license or within the operational restrictions indicated.

The removal of these conditions shall be made by an amendment to the renewed license supported by a favorable evaluation by the Commission.

2.C.(3)(a) Deleted per Amendment 24, 6/19/81.

Renewed License No. NPF-6 Amendment No. 307

REACTOR COOLANT SYSTEM STEAM GENERATOR (SG) TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 a. SG tube integrity shall be maintained, and

b. All SG tubes satisfying the tube plugging criteria shall be plugged in accordance with the Steam Generator Program.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

Note: ACTIONS may be entered separately for each SG tube.

a. With one or more SG tubes satisfying the tube plugging criteria and not plugged in accordance with the Steam Generator Program,

1. Within 7 days verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, and

2. Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following the next ~efueling outage or SG tube inspection.

b. If the required ACTION and Allowed Outage Time of ACTION a above cannot be met or SG tube integrity cannot be maintained, be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.5.1 Verify SG tube integrity in accordance with the Steam Generator Program.

4.4.5.2 Verify that each inspected SG tube that satisfies the tube plugging criteria is plugged in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a SG tube inspection.

ARKANSAS - UNIT 2 3/4 4-6 Amendment No. ~.4-8-7.~.U7.

Next Page is 3/4 4-13 ~.~.~.2W. 307

ADMINISTRATIVE CONTROLS 6.5.9 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm through any one SG.

3. The operational leakage performance criterion is specified in LCO 3.4.6.2, Reactor Coolant System Operational Leakage.

c. Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

ARKANSAS- UNIT 2 6-8 Amendment No. ~.29e, 307

ADMINISTRATIVE CONTROLS 6.5.9 Steam Generator (SG) Program (continued)

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.

2. After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; ARKANSAS - UNIT 2 6-9 Amendment No. ~.~, 307 Next Page is 6-14

ADMINISTRATIVE CONTROLS 6.5.9 Steam Generator (SG) Program (continued) c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.

3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary leakage.

6.5.10 Secondary Water Chemistry This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. The program shall include:

a. Identification of a sampling schedule for the critical variables and control points for these variables;

b. Identification of the procedures used to measure the values of the critical variables;

c. Identification of process sampling points;

d. Procedure for the recording and management of data;

e. Procedures defining corrective actions for all off control point chemistry conditions; and

f. A procedure identifying the authority responsible for the interpretation of the data, and the sequence and timing of administrative events required to initiate corrective action.

ARKANSAS - UNIT 2 6-14 Amendment No. ~. 307

ADMINISTRATIVE CONTROLS 6.6.6 Containment Inspection Report Any degradation exceeding the acceptance criteria of the containment structure detected during the tests required by the Containment Tendon Surveillance Program shall undergo an engineering evaluation within 60 days of the completion of the inspection surveillance. The results of the engineering evaluation shall be reported to the NRC within an additional 30 days of the time the evaluation is completed. The report shall include the cause of the condition that does not meet the acceptance criteria, the applicability of the conditions to the other unit, the acceptability of the concrete containment without repair of the item, whether or not repair or replacement is required and, if required, the extent, method, and completion date of necessary repairs, and the extent, nature, and frequency of additional examinations.

6.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed in accordance with the Specification 6.5.9, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each degradation mechanism,

f. The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG, and

g. The results of condition monitoring, including the results of tube pulls and in-situ testing.

6.6.8 Specific Activity The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded the results of one analysis after the radioiodine activity was reduced to less than limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

ARKANSAS - UNIT 2 6-22 Amendment No. ~,2-a-7,~,~, 307

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 307 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-6 ENTERGY OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNIT 2 DOCKET NO. 50-368

1.0 INTRODUCTION

By application dated April 24, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17114A399), Entergy Operations, Inc. (the licensee), requested changes to the technical specifications (TSs) for Arkansas Nuclear One, Unit 2 (AN0-2).

The proposed changes would revise the TSs for AN0-2, and would adopt U.S. Nuclear Regulatory Commission (NRC)-approved Technical Specification Task Force (TSTF) Standard Technical Specifications (STSs) Change Traveler TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection" (ADAMS Accession No. ML110610350). The guidance of TSTF-510 revises TS 3.4.17, "Steam Generator (SG)

Tube Integrity"; TS 5.5.9, "Steam Generator (SG) Program"; and TS 5.6.7, "Steam Generator Tube Inspection Report, of NUREG-1432, Revision 4, "Standard Technical Specifications -

Combustion Engineering Plants" (ADAMS Accession No. ML12102A165), applicable to AN0-2.

The specific changes concern SG inspection periods, and address applicable administrative changes and clarifications.

The licensee stated that the license amendment request (LAR) is consistent with the Notice of Availability of TSTF-510, Revision 2, announced in the Federal Registeron October 27, 2011 (76 FR 66763), as part of the consolidated line item improvement process.

The current STS requirements in the above specifications were established in May 2005 with the NRC staff's approval of TSTF-449, Revision 4, "Steam Generator Tube Integrity" (NRC Federal Register Notice of Availability (70 FR 24126)). The TSTF-449 changes to the STS incorporated a new, largely performance-based approach for ensuring the integrity of the SG tubes is maintained. The performance-based requirements were supplemented by prescriptive requirements relating to tube inspections and tube repair limits to ensure that conditions adverse to quality are detected and corrected on a timely basis. As of September 2007, the TSTF-449, Revision 4, changes were adopted in the plant TSs for all pressurized water reactors (PWRs).

Enclosure 2

The proposed changes in TSTF-510, Revision 2, reflect licensees' early implementation experience with respect to TSTF-449, Revision 4. TSTF-51 O characterizes the changes as editorial corrections, changes, and clarifications intended to improve internal consistency, consistency with implementing industry documents, and usability without changing the intent of the requirements. The proposed changes are an improvement to the existing SG inspection requirements and continue to provide assurance that the licensing basis will be maintained between SG inspections.

2.0 REGULATORY EVALUATION

The SG tubes in PW Rs have a number of important safety functions. These tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied upon to maintain primary system pressure and inventory. As part of the RCPB, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system and are relied upon to isolate the radioactive fission products in the primary coolant from the secondary system. In addition, the SG tubes are relied upon to maintain their integrity to be consistent with the containment objectives of preventing uncontrolled fission product release under conditions resulting from core damage during severe accidents.

The regulations in Title 1O of the Code of Federal Regulations (10 CFR) establish the requirements with respect to the integrity of SG tubing. Specifically, the General Design Criteria (GDC) in Appendix A to 10 CFR Part 50 state that the RCPB:

• shall have "an extremely low probability of abnormal leakage ... and gross rupture" (GDC 14, "Reactor pressure coolant boundary"),

• "shall be designed with sufficient margin" (GDC 15, "Reactor coolant system design,"

and GDC 31, "Fracture prevention of reactor coolant pressure boundary"),

• shall be of "the highest quality standards possible" (GDC 30, "Quality of reactor coolant pressure boundary"), and

• shall be designed to permit "periodic inspection and testing ... to assess ... structural and leaktight integrity" (GDC 32, "Inspection of reactor coolant pressure boundary").

The AN0-2 plant was designed and constructed to meet the intent of the Atomic Energy Commission's GDC, as originally proposed in July 1967, and thus, the design and construction were initiated and proceeded to a significant extent based upon the criteria proposed in 1967.

Section 3.1 of the AN0-2 Safety Analysis Report lists the manner in which the AN0-2 GDC meet the intent of the GDC in Appendix A of 10 CFR Part 50. The AN0-2 GDC addressing the RCPB are Criterion 14, "Reactor Coolant Pressure Boundary";

Criterion 15, "Reactor Coolant System Design"; Criterion 30, "Quality of Reactor Coolant Pressure Boundary"; Criterion 31, "Fracture Prevention of Reactor Coolant Pressure Boundary";

and Criterion 32, "Inspection of Reactor Coolant Pressure Boundary." These AN0-2 GDC are similar to GDC 14, 15, 30, 31, and 32 in Appendix A of 10 CFR Part 50.

The regulations in 10 CFR 50.55a specify that RCPB components must meet the requirements for Class 1 components in Section 111 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). Section 50.55a further requires, in part, that

throughout the service life of a PWR, ASME Code Class 1 components meet the requirements, except design and access provisions and preservice examination requirements, in Section XI, "Rules for lnservice Inspection (ISi) of Nuclear Power Plant Components," of the ASME Code, to the extent practical. This requirement includes the inspection and repair criteria of Section XI of the ASME Code.

As part of the plant's licensing basis, applicants for PWR licenses are required to analyze the consequences of postulated design-basis accidents such as an SG tube rupture or main steamline break. These analyses consider the primary-to-secondary leakage that may occur during these events and must show that the radiological consequences do not exceed the applicable limits of the 10 CFR Part 100.11, "Determination of exclusion area, low population zone, and population center distance," guidelines for offsite doses (or 10 CFR 50.67, "Accident source term," as appropriate), GDC-19 of Appendix A to 10 CFR 50 for control room operator doses (or some fraction thereof as appropriate to the accident), or the NRG-approved licensing basis.

The regulation at 10 CFR 50.36, "Technical specifications," establishes the requirements related to the content of the TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five categories related to station operation:

(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs);

(3) surveillance requirements (SRs);

(4) design features; and (5) administrative controls.

For AN0-2, the LCOs (and accompanying Action statements) and SRs in the TSs relevant to SG tube integrity are in TS 3.4.5, "Steam Generator (SG) Tube Integrity." The SRs in the "Steam Generator (SG) Tube Integrity" specification reference the SG Program, which is defined in the Administrative Controls section of the TSs.

The regulation at 10 CFR 50.36(c)(5) defines administrative controls as "the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure the operation of the facility in a safe manner." Programs established by the licensee to operate the facility in a safe manner, including the SG Program, are listed in the Administrative Controls section of the TSs. In the STSs, the SG Program is defined in TS 5.5.9 and the reporting requirements related to implementation of the SG Program are in TS 5.6.7.

TS 6.5.9, "Steam Generator (SG) Program," for AN0-2, requires that an SG Program be established and implemented to ensure that SG tube integrity is maintained. Tube integrity is maintained by meeting the performance criteria specified in TS 6.5.9.b for structural and leakage integrity, consistent with the plant design and licensing bases. TS 6.5.9.a requires that a condition monitoring assessment be performed during each outage, during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

TS 6.5.9.d includes provisions regarding the scope, frequency, and methods of SG tube inspections. These provisions require that the inspections be performed with the objective of detecting flaws of any type that (1) may be present along the length of a tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and (2) may satisfy the applicable tube plugging criteria.

The applicable tube repair criteria, specified in TS 6.5.9.c, are that tubes found during ISi to contain flaws with a depth equal to or exceeding 40 percent of the nominal wall thickness shall be plugged.

3.0 TECHNICAL EVALUATION

The changes proposed by the licensee for AN0-2 LCO 3.4.5, SR 4.4.5.2, TS 6.5.9, and TS 6.6. 7 are consistent with the corresponding TS changes described in TSTF-510, Revision 2, including the proposed revised inspection intervals, which are appropriate since AN0-2 has SGs with thermally treated Alloy 690 tubing. Except for the administrative changes and variations discussed below, the proposed TS changes are consistent with TSTF-510, Revision 2. As a result, the NRC staff's evaluation is focused on these administrative changes and variations, since the other changes were previously evaluated in the model safety evaluation (ADAMS Accession No. ML112101513), which is applicable to AN0-2.

3.1 Administrative Changes and Variations

• The AN0-2 TSs utilize different numbering than the STSs on which TSTF-51 O, Revision 2, was based. For AN0-2, the "Steam Generator (SG) Tube Integrity" TS is numbered 3.4.5 rather than 3.4.18.

• An NRC letter dated June 17, 2013 (ADAMS Accession No. ML13120A541) clarified that if LARs proposing to implement TSTF-510, Revision 2, corrected an administrative inconsistency in paragraph 5.5.9.d.2 (or 6.5.9.d.2 for AN0-2) of the Steam Generator (SG) Program, it would not result in removal of submitted LARs from the consolidated line item improvement process. Accordingly, since AN0-2 does not have any approved tube repair methods, this LAR fixes the administrative inconsistency in paragraph 6.5.9.d.2 by replacing "tube repair criteria" with "tube plugging criteria."

• As noted by the licensee in the LAR, the AN0-2 TSs are not based on the STS format, which results in administrative wording differences between the AN0-2 TSs and the STSs. For example, the AN0-2 TSs reference "Hot Shutdown" instead of "Mode 4" when referring to operational Mode 4. The AN0-2 TSs define the operational modes in Table1.1.

The differences noted above are administrative and do not affect the applicability of TSTF-510, Revision 2, to the AN0-2 TSs. As a result, the NRC staff finds that the differences between what was previously approved for TSTF-510, Revision 2, and the licensee's proposed TS changes, are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Arkansas State official was notified of the proposed issuance of the amendment on September 11, 2017. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes 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 published in the Federal Registeron July 5, 2017 (82 FR 31093).

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

6.0 CONCLUSION

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

Principal Contributor: A. Johnson Date: October 10, 2017

ML17251A211 *concurred via memo OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DSS/STSB/BC(A) NRR/DLR/RCCB/BC*

NAME TWengert PBlechman JWhitman SB loom DATE 9/15/17 9/11/17 9/19/17 07/17/2017 OFFICE OGC- NLO NRR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME JGillespie RPascarelli TWengert DATE 10/3/17 10/6/17 10/10/17