ML16165A423

Issuance of Amendment No. 257 Adoption of Technical Specifications Task Force Traveler TSTF-545, Revision 3,

ML16165A423
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/10/2017
From:	Thomas Wengert Plant Licensing Branch IV
To:	Entergy Operations
Koenick S
References
CAC MF7536
Download: ML16165A423 (25)

Text

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

1448 S.R. 333 Russellville, AR 72802

SUBJECT:

ARKANSAS NUCLEAR ONE, UNIT 1 - ISSUANCE OF AMENDMENT RE: ADOPTION OF TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER TSTF-545, REVISION 3, TS INSERVICE TESTING PROGRAM REMOVAL &

CLARIFY SR USAGE RULE APPLICATION TO SECTION 5.5 TESTING" (CAC NO. MF7536)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 257 to Renewed Facility Operating License No. DPR-51 for Arkansas Nuclear One, Unit 1. The amendment consists of changes to the technical specifications (TS) in response to your application dated March 25, 2016.

The amendment deletes TS 5.5.8, "lnservice Testing Program." A new defined term, "lnservice Testing Program," is added to TS Section 1.1, "Definitions." In addition, existing uses of the term "lnservice Testing Program" in the TSs are capitalized throughout to indicate that it is now a defined term. These changes are based on NRC-approved Technical Specifications Task Force (TSTF) Standard Technical Specifications Change Traveler TSTF-545, Revision 3, TS lnservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing."

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. Wengert, Senior roject Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-313

Enclosures:

1. Amendment No. 257 to DPR-51

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

ML16165A423 *via e-mail OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DSS/STSB/BC*

NAME TWengert PBlechman AKlein DATE 03/01/17 03/01/17 03/01/17 *.*

OFFICE DE/EPNB/BC OGG NLO w/comments NRR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME DAiiey DRoth RPascarelli TWengert DATE 02/28/17 03/07/17 03/10/17 03/10/17 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY OPERATIONS, INC.

DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE. UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 257 Renewed License No. DPR-51

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

A. The application for amendment by Entergy Operations, Inc. (the licensee), dated March 25, 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 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. DPR-51 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 257, are hereby incorporated in the renewed license. EOI 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. DPR-51 and Technical Specifications Date of Issuance: March 1 o, 2o1 7

ATTACHMENT TO LICENSE AMENDMENT NO. 257 RENEWED FACILITY OPERATING LICENSE NO. DPR-51 ARKANSAS NUCLEAR ONE. UNIT 1 DOCKET NO. 50-313 Replace the following pages of the Renewed Facility Operating License No. DPR-51 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 1.1-3 1.1-3 3.4.10-2 3.4.10-2 3.4.14-2 3.4.14-2 3.5.2-1 3.5.2-1 3.6.3-4 3.6.3-4 3.6.5-3 3.6.5-3 3.7.1-2 3.7.1-2 3.7.1-3 3.7.2-2 3.7.2-2 3.7.3-2 3.7.3-2 3.7.5-2 3.7.5-2 5.0-11 5.0-11

(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 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; (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 the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 2568 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 257, are hereby incorporated in the renewed license.

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

(3) Safety Analysis Report The licensee's SAR supplement submitted pursuant to 10 CFR 54.21(d),

as revised on March 14, 2001, describes certain future inspection activities to be completed before the period of extended operation. The licensee shall complete these activities no later than May 20, 2014.

(4) Physical Protection EOI 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 combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Arkansas Nuclear One Physical Security Plan, Training and Qualifications Plan, and Safeguards Contingency Plan," as submitted on May 4, 2006.

Renewed License No. DPR-51 Amendment No. 257 Revised by letter dated July 18, 2007

Definitions 1.1 1.1 Definition (continued)

DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except RCP seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection and leakoff) that is not identified LEAKAGE;

c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

AN0-1 1.1-3 Amendment No. ~.224.~. 257

Pressurizer Safety Valves 3.4.10 CONDITION REQUIRED ACTION COMPLETION TIME C. Required pressurizer C.1 Be in MODE 4 with RCS 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> safety valve inoperable in temperature ~ 259 °F.

MODE 3 or MODE 4 with RCS temperature> 259 °F.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each required pressurizer safety valve is In accordance with OPERABLE in accordance with the INSERVICE the INSERVICE TESTING PROGRAM. Following testing, as-left TESTING PROGRAM lift settings shall be within +/- 1%.

AN0-1 3.4.10-2 Amendment No. ~.2-M,257

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 ----------------------------N 0 TE---------------------------------

Not required to be performed in MODES 3 and 4.

Verify leakage from each RCS pressure isolation In accordance with check valve, or pair of check valves, as applicable, is the INSERVICE less than or equal to an equivalent of the Allowable TESTING Leakage Limit identified below at a differential test PROGRAM pressure z 150 psid.

Pressure Isolation Allowable Check Valve(s) Leakage Limit Once prior to entering MODE 2 DH-14A ~5 gpm whenever the unit DH-13A and DH-17 ~ 5 gpm total has been in DH-148 ~5 gpm MODE 5 for 7 days DH-138 and DH-18 ~ 5 gpm total or more, if leakage testing has not been performed in the previous 9 months SR 3.4.14.2 Verify OHR System autoclosure interlock prevents 18 months the valves from being opened with a simulated or actual high RCS pressure signal.

SR 3.4.14.3 Verify OHR System autoclosure interlock causes the 18 months valves to close automatically with a simulated or actual high RCS pressure signal:

c. ~ 340 psig for one valve; and

d. ~ 400 psig for the other valve.

SR 3.4.14.4 Verify OHR System autoclosure interlock prevents 18 months the valves from being opened with a simulated or actual Core Flood Tank isolation valve "not closed" signal.

SR 3.4.14.5 Verify OHR System autoclosure interlock causes the 18 months valves to close automatically with a simulated or actual Core Flood Tank isolation valve "not closed" signal.

AN0-1 3.4.14-2 Amendment No.~. 257

ECCS - Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, MODE 3 with Reactor Coolant System (RCS) temperature> 350 °F.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains A.1 Restore train(s) to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Reduce RCS temperature 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to s 350 °F.

C. Less than 100% of the C.1 Enter LCO 3.0.3. Immediately ECCS flow equivalent to a single OPERABLE train available.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify each ECCS manual, power operated, and 31 days automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.2.2 Verify each ECCS pump's developed head at the In accordance with test flow point is greater than or equal to the the INSERVICE required developed head. TESTING PROGRAM AN0-1 3.5.2-1 Amendment No. ~. 257

Reactor Building Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1 Verify each reactor building purge isolation valve is 31 days closed.

SR 3.6.3.2 -------------------------------N 0 TE-----------------------------

Va Ives and blind flanges in high radiation areas may be verified by use of administrative means.

Verify each reactor building isolation manual valve 31 days and blind flange that is located outside the reactor building and not locked, sealed, or otherwise secured, and is required to be closed during accident conditions is closed, except for reactor building isolation valves that are open under administrative controls.

SR 3.6.3.3 -------------------------------N 0 TE-----------------------------

Va Ives and blind flanges in high radiation areas may be verified by use of administrative means.

Verify each reactor building isolation manual valve Prior to entering and blind flange that is located inside the reactor MODE 4 from building and not locked, sealed, or otherwise MODE 5 if not secured, and required to be closed during accident performed within conditions is closed, except for reactor building the previous isolation valves that are open under administrative 92 days controls.

SR 3.6.3.4 Verify the isolation time of each automatic power In accordance with operated reactor building isolation valve is within the INSERVICE limits. TESTING PROGRAM SR 3.6.3.5 Verify each automatic reactor building isolation 18 months valve that is not locked, sealed, or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.

AN0-1 3.6.3-4 Amendment No. ~.~.257

Reactor Building Spray and Cooling System 3.6.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify each reactor building spray manual, power 31 days operated, and automatic valve in each required flow path that is not locked, sealed, or otherwise secured in position is in the correct position.

SR 3.6.5.2 Operate each required reactor building cooling train 31 days fan unit for ~ 15 minutes.

SR 3.6.5.3 Verify each required reactor building cooling train 31 days cooling water flow rate is ~ 1200 gpm.

SR 3.6.5.4 Verify each required reactor building spray pump's In accordance with developed head at the flow test point is greater than the INSERVICE or equal to the required developed head. TESTING PROGRAM SR 3.6.5.5 Verify each automatic reactor building spray valve in 18 months each required flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.6.5.6 Verify each required reactor building spray pump 18 months starts automatically on an actual or simulated actuation signal.

SR 3.6.5.7 Verify each required reactor building cooling train 18 months starts automatically on an actual or simulated actuation signal.

SR 3.6.5.8 Verify each spray nozzle is unobstructed. Following maintenance which could result in nozzle blockage AN0-1 3.6.5-3 Amendment No. ~.~. 257

MSSVs 3.7.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 ---------------------------------N 0 TE----------------------------

0 nIy required to be performed in MODES 1 and 2.

Verify each required MSSV lift setpoint in In accordance with accordance with the INSERVICE TESTING the INSERVICE PROGRAM. Following testing, as-left lift settings TESTING shall be within +/- 1 %. PROGRAM Table 3.7.1-1 Allowable Power Level and RPS Nuclear Overpower Trip Allowable Value versus OPERABLE Main Steam Safety Valves RPS NUCLEAR MINIMUM NUMBER OF MAXIMUM ALLOWABLE OVERPOWER TRIP MSSVS OPERABLE POWER LEVEL ALLOWABLE VALUE (PER SG) (% RTP)

(% RTP) 6 85.7 89.9 5 71.4 74.9 4 57.1 59.9 3 42.8 44.9 2 28.5 29.9 AN0-1 3.7.1-2 Amendment No. 2+a, 257

MS IVs 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 --------------------------------N 0 TE-----------------------------

0 nIy required to be performed in MODES 1 and 2.

Verify isolation time of each MSIV is within the limits In accordance with specified in the INSERVICE TESTING PROGRAM. the INSERVICE TESTING PROGRAM SR 3.7.2.2 --------------------------------N 0 TE-----------------------------

1. Only required to be performed in MODES 1 and 2.

2. Not required to be met when SG pressure is

< 750 psig.

Verify each MSIV actuates to the isolation position 18 months on an actual or simulated actuation signal.

AN0-1 3.7.2-2 Amendment No. ~. 257

MFIVs, Main Feedwater Block Valves, Low Load Feedwater Control Valves and Startup Feedwater Control Valves 3.7.3 CONDITION REQUIRED ACTION COMPLETION TIME D. One Startup Feedwater D.1 Close or isolate Startup 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Control Valve in one or Feedwater Control Valve.

more flow paths inoperable. AND D.2 Verify Startup Feedwater Once per 7 days Control Valve is closed or isolated.

E. Two valves in the same E.1 Isolate affected flow path. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> flow path inoperable for one or more flow paths.

F. Required Action and F.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND F.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 --------------------------------N 0 TE-----------------------------

0 n ly required to be performed in MODES 1 and 2.

Verify the isolation time of each MFIV, Main In accordance with Feedwater Block Valve, Low Load Feedwater the INSERVICE Control Valve and Startup Feedwater Control Valve TESTING is within the limits provided in the INSERVICE PROGRAM TESTING PROGRAM.

AN0-1 3.7.3-2 Amendment No. ~. 257

EFW System 3.7.5 CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or B AND not met.

C.2 Be in MODE 4. 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> D. Two EFW trains D.1 -------------NOTE----------------

inoperable in MODE 1, 2, LCO 3.0.3 and all other or 3. LCO Required Actions requiring MODE changes are suspended until one EFW train is restored to OPERABLE status.

Initiate action to restore Immediately one EFW train to OPERABLE status.

E. Required EFW train E.1 Initiate action to restore Immediately inoperable in MODE 4. EFW train to OPERABLE status.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify each EFW manual, power operated, and 31 days automatic valve in each water flow path and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.7.5.2 -------------------------------N 0 TE------------------------------

Not required to be performed for the turbine driven EFW pump, until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching ~ 750 psig in the steam generators.

Verify the developed head of each EFW pump at In accordance with the flow test point is greater than or equal to the the INSERVICE required developed head. TESTING PROGRAM AN0-1 3.7.5-2 Amendment No. ~. 257

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.7 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel. Surface and volumetric examination of the reactor coolant pump flywheels will be conducted coincident with refueling or maintenance shutdowns such that during 1O year intervals all four reactor coolant pump flywheels will be examined. Such examinations will be performed to the extent possible through the access ports, i.e., those areas of the flywheel accessible without motor disassembly. The surface and volumetric examination may be accomplished by Acoustic Emission Examination as an initial examination method. Should the results of the Acoustic Emission Examination indicate that additional examination is necessary to ensure the structural integrity of the flywheel, then other appropriate NOE methods will be performed on the area of concern.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program inspection frequencies.

5.5.8 DELETED AN0-1 5.0-11 Amendment No. ~.2-JQ.,~.~. 257

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 257 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-51 ENTERGY OPERATIONS. INC.

ARKANSAS NUCLEAR ONE, UNIT 1 DOCKET NO. 50-313

1.0 INTRODUCTION

By application dated March 25, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16088A181), Entergy Operations, Inc. (Entergy, the licensee), requested changes to the technical specifications (TSs) for Arkansas Nuclear One, Unit 1 (AN0-1). Specifically, the licensee requested changes to the TSs consistent with Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-545, Revision 3, "TS lnservice Testing Program Removal & Clarify SR (Surveillance Requirement] Usage Rule Application to Section 5.5 Testing, dated October 21, 2015 (ADAMS Accession No. ML15294A555).

The licensee's proposed changes delete AN0-1 TS 5.5.8, "lnservice Testing Program," and adds a new defined term, "INSERVICE TESTING PROGRAM," to the TSs. All existing references to the "lnservice Testing Program" in the AN0-1 TS SRs are replaced with "INSERVICE TESTING PROGRAM" so that the SRs refer to the new definition in lieu of the deleted program. The licensee also proposed additional TS format changes unrelated to TSTF-545, Revision 3.

The licensee's letter dated March 25, 2016, also included a request to use American Society of Mechanical Engineers (ASME) Code Case OMN-20, "lnservice Test Frequency," as an alternative to certain ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) requirements at AN0-1. The U.S. Nuclear Regulatory Commission (NRC) considered this request separately from the proposed license amendment, and authorized the licensee's use of this alternative by letter dated December 9, 2016 (ADAMS Accession No. ML16130A471).

Enclosure 2

2.0 REGULATORYEVALUATION

2.1 Description of lnservice Testing Requirements and TSTF-545 An inservice test is a test to assess the operational readiness of a structure, system, or component after first electrical generation by nuclear heat. The ASME OM Code provides requirements for inservice testing of certain components in light-water nuclear power plants. The ASME OM Code identifies the components subject to the testing (i.e., pumps, valves, pressure relief devices, and dynamic restraints), responsibilities, methods, intervals, parameters to be measured and evaluated, criteria for evaluating results, corrective actions, personnel qualification, and recordkeeping. Title 10 of the Code of Federal Regulations (10 CFR), paragraph 50.55a(f), "lnservice testing requirements," requires that inservice testing of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda. The facility's TSs also prescribe inservice testing requirements and frequencies for ASME Code Class 1, 2, and 3 components.

The regulation in 10 CFR 50.55a(f)(5)(ii) states, in part, "If a revised inservice test program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program." TSTF-545, Revision 3, provides guidance to licensees on how to request license amendments that would eliminate conflicting requirements between 10 CFR 50.55a and the TSs. TSTF-545, Revision 3, proposes elimination of the lnservice Testing Program from the Administrative Controls section of the TSs. The TSs contain surveillances that require testing or test intervals in accordance with the lnservice Testing Program. The elimination of the lnservice Testing Program from the TSs could cause uncertainty regarding the correct application of these SRs. Therefore, TSTF-545, Revision 3, also proposes adding a new definition, "INSERVICE TESTING PROGRAM," to the TSs, which would be defined as "the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." TSTF-545, Revision 3, proposes replacement of existing uses of the term, "lnservice Testing Program," with the defined term, as denoted by capitalized letters, throughout the TSs.

The NRC approved TSTF-545, Revision 3, by letter dated December 11, 2015 (ADAMS Package Accession No. ML15317A071), and published a notice of availability in the Federal Register (FR) on March 28, 2016 (81 FR 17208).

2.2 Proposed Technical Specifications Changes The licensee requested to delete TS 5.5.8 from the Administrative Controls section of TSs and replace it with the word "DELETED." TS 5.5.8 currently states:

This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a. Testing frequencies applicable to the ASME Code for Operation and Maintenance (OM) of Nuclear Power Plants and applicable Addenda as follows:

ASME OM Code terminology for Required Frequencies for performing inservice testing activities inservice testing activities Monthly At least once per 31 days Every 6 weeks At least once per 42 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the lnservice Testing Program for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

SR 3.0.2 allows an extension of inservice testing intervals by up to 25 percent. If it is discovered that a surveillance associated with an inservice testing activity was not performed within the required interval, SR 3.0.3 allows the licensee to delay declaring the associated limiting condition for operation not met in order to perform the missed surveillance. The licensee did not request changes to SR 3.0.2 or SR 3.0.3.

The licensee requested to revise the Definitions section of TSs by adding the term, "INSERVICE TESTING PROGRAM," with the following definition: "The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." The licensee also requested that all existing occurrences of "lnservice Testing Program" in TS SRs be replaced with "INSERVICE TESTING PROGRAM," so that the SRs refer to the new definition in lieu of the deleted program.

The licensee requested to move TS Table 3.7.1-1, in its entirety, from TS page 3.7.1-3 to TS page 3.7.1-2, and to delete TS page 3.7.1-3. Currently, TS page 3.7.1-3 does not include any requirements other than Table 3.7.1-1.

2.3 Regulatory Requirements and Guidance The NRC staff considered the following regulatory requirements, guidance, and licensing information during its review of the proposed changes:

Technical Specifications Paragraph 50.36(c) of 10 CFR requires TSs to include the following categories: (1) safety limits, limiting safety systems settings, and control settings; (2) limiting conditions for operation; (3) SRs; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports. Section 50.36(c)(3) of 10 CFR states that "[s]urveillance requirements are requirements relating to test, calibration, or inspection to assure that the

necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." Section 50.36(c)(5) of 10 CFR states that "[a]dministrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner."

The NRC staff's guidance for review of the TSs is in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]

Edition," Chapter 16, "Technical Specifications," Revision 3, dated March 2010 (ADAMS Accession No. ML100351425). As described therein, as part of the regulatory standardization effort, the staff has prepared improved STSs for each of the LWR nuclear steam supply systems and associated balance-of-plant equipment systems. In April of 2012, the NRC most-recently published NUREG-1430, Revision 4, "Standard Technical Specifications - Babcock and Wilcox Plants" (ADAMS Accession No. ML 121OOA177).

The licensee's proposed amendment is based on TSTF-545, Revision 3, which is an NRC-approved change to the improved STSs. The staff's review includes consideration of whether the proposed changes are consistent with TSTF-545, Revision 3. Special attention is given to TS provisions that depart from the improved STSs, as modified by NRG-approved TSTF travelers, to determine whether proposed differences are justified by uniqueness in plant design or other considerations so that 10 CFR 50.36 is met. In addition, the guidance states that comparing the change to previous STS can help clarify the TS intent.

lnservice Testing Pursuant to 10 CFR 50.54, "Conditions of licenses," the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include inservice testing of pumps and valves at nuclear power reactors in accordance with the ASME OM Code as specified in 10 CFR 50.55a(f). The regulations in 10 CFR 50.55a(f) state, in part:

Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV [Boiler and Pressure Vessel] Code and ASME Code for Operation and Maintenance of Nuclear Power Plants as specified in this paragraph. Each operating license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions

[referring to 10 CFR 50.55a(f)(1) through (f)(6)] ....

The ASME OM Code is a consensus standard, which is incorporated by reference into 10 CFR 50.55a. During the incorporation process, the NRC staff reviewed the ASME OM Code requirements for technical sufficiency and found that the ASME OM Code inservice testing program requirements were suitable for incorporation into the NRC's rules.

The regulation in 10 CFR 50.55(a)(f)(5)(ii) states, in part: "If a revised inservice test program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program."

NUREG-1482, Revision 2, "Guidelines for lnservice Testing at Nuclear Power Plants: lnservice Testing of Pumps and Valves and lnservice Examination and Testing of Dynamic Restraints

(Snubbers) at Nuclear Power Plants," Final Report, October 2013 (ADAMS Accession No. ML13295A020) provides guidance for the inservice testing of pumps and valves.

NUREG-0800, Section 3.9.6, "Functional Design, Qualification, and lnservice Testing Programs for Pumps, Valves, and Dynamic Restraints," Revision 3, March 2007 (ADAMS Accession No. ML070720041 ), provides guidance and acceptance criteria for the NRC staff review of the inservice testing program for pumps and valves.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the licensee's application to determine if the proposed changes are consistent with the guidance, regulations, and licensing information discussed in Section 2.3 of this SE. In determining whether an amendment to a license will be issued, the Commission is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Among the considerations are whether the TSs, as amended, would provide the necessary administrative controls per 10 CFR 50.36(c)(5) (i.e., provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner). In making its determination as to whether to amend the license, the staff considered those regulatory requirements that are automatically conditions of the license through 10 CFR 50.54. Where the regulations already condition the license, there is no need for a duplicative requirement in the TSs; the regulations provide the necessary reasonable assurance of the health and safety of the public.

3.1 Deletion of the lnservice Testing Program from the TSs TS 5.5.8 requires the licensee to have an inservice testing program that provides controls for inservice testing of ASME Code Class 1, 2, and 3 components (i.e., pumps and valves).

Through 10 CFR 50.54, the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include 10 CFR 50.55a(f), which specifies the requirements for the inservice testing of pumps and valves. Therefore, requiring the licensee to have an inservice testing program in TSs is duplicative of the license condition in 10 CFR 50.54. Thus, with the proposed TS changes, the licensee will still be required to maintain an inservice testing program in accordance with the ASME OM Code, as specified in 10 CFR 50.55a(f). For the reasons explained below, it is not necessary to have additional administrative controls in the TSs relating to the inservice testing program to assure operation of the facility in a safe manner.

Consideration of TS 5. 5. 8. a The ASME OM Code requires testing to normally be performed within certain time periods.

TS 5.5.8.a sets inservice testing frequencies more precisely than those specified in the ASME OM Code and applicable addenda (e.g., "at least once per 31 days" contrasted with "monthly").

However, the NRC staff determined that the more precise inservice testing frequencies are not necessary to assure operation of the facility in a safe manner.

Consideration of TS 5.5.8.b TS 5.5.8.b allows the licensee to extend, by up to 25 percent, the interval between inservice testing activities, as required by TS 5.5.8.a and for other normal and accelerated frequencies specified as 2 years or less in the inservice testing program. Similar to TS 5.5.8.b, the NRC

authorization of Code Case OMN-20 on December 9, 2016, also permits the licensee to extend the inservice testing intervals specified in the ASME OM Code by up to 25 percent.

The NRC staff determined that the TS 5.5.8.b allowance to extend inservice testing intervals is not needed to assure operation of the facility in a safe manner. Therefore, the NRC staff determined that deletion of TS 5.5.8.b is acceptable. The deletion of TS 5.5.8.b does not impact the licensee's ability to extend inservice testing intervals using Code Case OMN-20, as authorized by the NRC.

Consideration of TS 5. 5. 8. c TS 5.5.8.c allows the licensee to use SR 3.0.3 when it discovers that an SR associated with an inservice test was not performed within its specified frequency. SR 3.0.3 allows the licensee to delay declaring a limiting condition for operation not met in order to perform the missed surveillance. The use of SR 3.0.3 for inservice tests is limited to those inservice tests required by an SR. In accordance with 10 CFR 50.55a, the licensee may also request relief from the ASME OM Code requirements to address issues associated with a missed inservice test.

Deletion of TS 5.5.8.c does not change any of these requirements, and SR 3.0.3 will continue to apply to those inservice tests required by SRs. Based on the above, the NRC staff determined that deletion of TS 5.5.8.c is acceptable.

Consideration of TS 5. 5. 8. d TS 5.5.8.d states that nothing in the ASME OM Code shall be construed to supersede the requirements of any TS. However, the regulations in 10 CFR 50.55a(f)(5)(ii) address what to do if a revised inservice testing program for a facility conflicts with the TSs for the facility; they require the licensee to apply for an amendment to the TSs to conform the TSs to the revised program at least 6 months prior to the start of the period for which the provisions become applicable. Accordingly, there is no need for a TS stating how to address conflicts between the TSs and the inservice testing program because the regulations specify how conflicts must be resolved.

Conclusion Regarding Deletion of TS 5. 5. 8 The NRC staff determined that the requirements currently in TS 5.5.8 are not necessary to assure operation of the facility in a safe manner. Based on this evaluation, the staff concludes that deletion of TS 5.5.8 from the licensee's TSs is acceptable, because TS 5.5.8 is not required by 10 CFR 50.36(c)(5).

3.2 Definition of INSERVICE TESTING PROGRAM and Revision to SRs The licensee proposes to revise the TS Definitions section to include the term, "INSERVICE TESTING PROGRAM," with the following definition: "The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." The proposed definition of the INSERVICE TESTING PROGRAM is consistent with the definition in TSTF-545, Revision 3. The definition is acceptable to the NRC staff because it correctly refers to the inservice testing requirements in 10 CFR 50.55a(f).

The licensee requested that all existing references to the "lnservice Testing Program" in SRs be revised to "INSERVICE TESTING PROGRAM" to reference the new TS defined term in lieu of the deleted program. The proposed change is consistent with the intent of TSTF-545,

Revision 3, to replace the current references in SRs with the new definition. The NRC staff verified that for each SR reference to the "lnservice Testing Program," the licensee proposed to change the reference to "INSERVICE TESTING PROGRAM." The proposed change does not alter how the SR testing is performed. However, the inservice testing frequencies could change because the TSs will no longer include the more precise test frequencies in TS 5.5.8.a. As discussed in Section 3.1 of this SE, the staff determined that the TSs do not need to include the more precise testing frequencies currently in TS 5.5.8.a. Based on its review, the staff determined that revising the SRs to refer to the new definition is acceptable because these SRs will continue to be performed in accordance with the requirements of 10 CFR 50.55a(f). The staff also determined that, with the proposed changes that allow less-precise testing frequencies, 10 CFR 50.36(c)(3) will continue to be met because the SRs will continue to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

3.3 Deviations from TSTF-545 In its application, the licensee identified the following deviations from TSTF-545, Revision 3:

1. TSTF-545, Revision 3, completely deletes TS 5.5.8 from the TSs and renumbers the subsequent TS programs. The licensee proposes to delete the content of TS 5.5.8, but retains the TS number, and adds the word "DELETED." The licensee did not propose to renumber the subsequent TS programs.

2. Some of the numbering and wording for SRs that are modified does not match TSTF-545, Revision 3. However, the licensee stated that the SRs are equivalent.

The NRC staff finds that the proposed deviations are editorial in nature and the licensee's proposed TS changes remain consistent with the intent of TSTF-545, Revision 3. Therefore, the staff finds that the licensee's proposed TS changes are acceptable.

3.4 TS Table 3.7.1-1 The licensee proposes to move TS Table 3.7.1-1 in its entirety from TS page 3.7.1-3 to TS page 3.7.1-2. In addition, as Table 3.7.1-1 is the only text on TS page 3.7.1-3, the licensee requested that this page be deleted. The NRC staff reviewed these changes and concluded that they are formatting changes that do not alter any TS requirements, and therefore, 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 January 17, 2017. The State official had no comments.

5.0 ENVIRONMENTALCONSIDERATION The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes SRs.

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 FR on June 7, 2016 (81 FR 36619). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

Principal Contributors: Blake Purnell, NRR Caroline Tilton, NRR John Huang, NRR Date: March 10, 2017