ML17144A082

Issuance of Amendment No. 259 Adoption of TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing

ML17144A082
Person / Time
Site:	Cooper
Issue date:	06/20/2017
From:	Thomas Wengert Plant Licensing Branch IV
To:	Higginbotham K Nebraska Public Power District (NPPD)
Wengert T, NRR/DORL/LPLIV, 415-4037
References
CAC MF8321
Download: ML17144A082 (14)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 20, 2017 Mr. Kenneth Higginbotham Vice President-Nuclear and CNO Nebraska Public Power District 72676 648A Avenue Brownville, NE 68321

SUBJECT:

COOPER NUCLEAR STATION - 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. MF8321)

Dear Mr. Higginbotham:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 259 to Renewed Facility Operating License No. DPR-46 for the Cooper Nuclear Station (CNS). The amendment consists of changes to the technical specifications (TSs) in response to your application dated August 26, 2016 (Agencywide Documents Access and Manage.ment System (ADAMS) Accession No. ML16245A288).

The amendment deletes TS 5.5.6, "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. The NRC staff has concluded that the proposed amendment is consistent with 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," which was made available to the TSTF via an NRC letter dated December 11, 2015 (ADAMS Accession No. ML15317A071), as part of the consolidated line item improvement process.

The use of Code Case OMN-20, "lnservice Test Frequency," was previously approved for use at CNS in an NRC safety evaluation dated February 12, 2016 (ADAMS Accession No. ML16014A174).

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

Sincerely, 4fi'~&~

Thomas J. Wengert, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-298

Enclosures:

1. Amendment No. 259 to DPR-46

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 259 Renewed License No. DPR-46

1. The U.S. Nuclear Regulatory Commission (the Commission) has found th'at:

A. The application for amendment by Nebraska Public Power District (the licensee),

dated August 26, 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-46 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 259, are hereby incorporated in the 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 60 days from the date of issuance.

FOR THE NUCLEAR REGULA TORY 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-46 and Technical Specifications Date of Issuance: June 20, 201 7

ATTACHMENT TO LICENSE AMENDMENT NO. 259 RENEWED FACILITY OPERATING LICENSE NO. DPR-46 COOPER NUCLEAR STATION DOCKET NO. 50-298 Replace the following page of the Renewed Facility Operating License No. DPR-46 and the Appendix A, Technical Specifications with the enclosed revised pages. The revised pages are identif.ied by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility Operating License REMOVE INSERT Technical Specifications REMOVE INSERT 1.1-3 1.1-3 1.1-4 1.1-4 1.1-5 1.1-5 3.1-22 3.1-22 3.4-7 3.4-7 3.5-5 3.5-5 3.5-10 3.5-10 3.6-13 3.6-13 3.6-14 3.6-14 3.6-26 3.6-26 3.6-32 3.6-32 3.6-39 3.6-39 5.0-10 5.0-10

(5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

C. This 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 The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2419 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 259, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Physical Protection 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 combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Cooper Nuclear Station Safeguards Plan," submitted by letter dated May 17, 2006.

NPPD 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 NPPD CSP was approved by License Amendment No. 238 as supplemented by changes approved by License Amendments 244 and 249.

(4) Fire Protection NPPD shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the license amendment request dated April 24, 2012 (and supplements dated July 12, 2012, January 14, 2013, February 12, 2013, March 13, 2013, June 13, 2013, December 12, 2013, January 17, 2014, February 18, 2014, and April 11, 2014), and as approved in the safety evaluation dated April 29, 2014.

Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if Amendment No. 259

Definitions 1.1 1.1 Definitions DOSE EQUIVALENT 1-131 1-133, 1-134, and 1-135 actually present. The DOSE (continued) EQUIVALENT 1-131 concentration is calculated as follows:

DOSE EQUIVALENT 1-131=(1-131)+0.0060 (1-132) + 0.17 (1-133) + 0.0010 (1-134) + 0.029 (1-135). The dose conversion factors used for this calculation are those listed in Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.

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

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE into the drywall, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2. LEAKAGE into the drywall 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;

b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE;

d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per unit length of RATE (LHGR) fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

(continued)

Cooper 1.1-3 Amendment No. 259

Definitions 1.1 1.1 Definitions LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all TEST logic components required for OPERABILITY of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power ratio (CPR)

RATIO (MCPR) that exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE - OPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including REPORT (PTLR) heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 2419 MWt.

REACTOR PROTECTION The RPS RESPONSE TIME shall be that time segment from SYSTEM(RPS)RESPONSE the time the sensor contacts actuate to the time the scram TIME solenoid valves deenergize.

SHUTDOWN MARGIN (SOM) SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

(continued)

Cooper 1.1-4 Amendment No. 259

Definitions 1.1 1.1 Definitions SHUTDOWN MARGIN (SOM) a. The reactor is xenon free; (continued)

b. The moderator temperature is~ 68°F, corresponding to the most reactive state; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SOM.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TURBINE BYPASS SYSTEM The TURBINE BYPASS SYSTEM RESPONSE TIME RESPONSE TIME consists of two components:

a. The time from initial movement of the main turbine stop valve or control valve until 80% of the turbine bypass capacity is established; and

b. The time from initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

Cooper 1.1-5 Amendment No. 259 I

SLC System 3.1.7

.. S~_RVEILLANC~ R.f=OUIREMENTS (continued)

SURVEILLANCE FREQUENCY (continued) Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored within the limits of Figure 3.1.7-2 SR 3.1.7.6 Verify each SLC subsystem manual valve in the flow In accordance with path that is not locked, sealed, or otherwise secured the Surveillance in position, is in the correct position or can be aligned Frequency Control to the correct position. Program SR 3.1.7.7 Verify each pump develops a flow rate ~ 38.2 gpm at In accordance a discharge pressure ~ 1300 psig. with the INSERVICE TESTING PROGRAM SR 3.1.7.8 Verify flow through' one SLC subsystem from pump In accordance with into reactor pressure vessel. the Surveillance Frequency Control Program SR 3.1.7.9 Verify all heat traced piping between storage tank In accordance with and pump suction is unblocked. the Surveillance Frequency Control Program Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored within the limits of Figure 3.1.7-2 Cooper 3.1-22 Amendment No. 259

SRVs and SVs 3.4.3 SURVEILLANCE REQUIREMENTS

• = ~ *-*--~~-~~

FREQUENCY

___ _______SURVEILLANCE SR 3.4.3.1 Verify the safety function lift setpoints of the SRVs In accordance with and SVs are as follows: the INSERVICE TESTING PROGRAM Number of Setpoint SRVs _ __Jps~g)___

2 1080 +/- 32.4 3 1090 +/- 32.7 3 1100 +/- 33.0 Number of Setpoint SVs ---*---~1R~l91_ ' ' '

3 1240 +/- 37.2 Following testing, lift settings shall be within+/- 1%.

SR 3.4.3.2 ---------------------------NOTE------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify each SRV opens when manually actuated. In accordance with the Surveillance Frequency Control Program Cooper 3.4-7 Amendment No. 259

ECCS - Operating 3.5.1 SURVEILLAN~E::, ~§_Qlj l~_~M~N_T~J~~t_in_u_ed~)_ __

SURVEILLANCE FREQUENCY SR 3.5.1.6 Verify the following ECCS pumps develop the In accordance specified flow rate against a system head with the corresponding to the specified reactor pressure. INSERVICE TESTING SYSTEM HEAD PROGRAM NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF Core Spray <?: 4720 gpm 1 <?: 113 psig LPCI <?: 15,000 gpm 2 <?: 20 psig SR 3.5.1.7 ---------------------------NOTE--------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure s 1020 and ~ 920 psig, In accordance with the HPCI pump can develop a flow rate <?: 4250 gpm the Surveillance against a system head corresponding to reactor Control pressure.

SR 3.5.1.8 ------------~-~-------1\JOTE-------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressures 165 psig, the HPCI In accordance with pump can develop a flow rate ~ 4250 gpm against a the Surveillance system head corresponding to reactor pressure. Frequency Control Program (continued)

Cooper 3.5-5 Amendment No. 259

ECCS - Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

In-*********-******

SURVEILLANCE I FREQUENCY 1

SR 3.5.2.4 Verify each required ECCS pump develops the In accordance specified flow rate against a system head with the corresponding to the specified reactor pressure. INSERVICE TESTING SYSTEM HEAD PROGRAM NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF cs ;a:; 4720 gpm 1 ;a:; 113 psig LPCI ;a:; 7700 gpm 1 ~ 20 psig SR 3.5.2.5 -------------------------------NOTE-------------------------------

Vessel injection/spray may be excluded.

Verify each required ECCS injection/spray subsystem In accordance with actuates on an actual or simulated automatic initiation the Surveillance signal. Frequency Control Program Cooper 3.5-10 Amendment No. 259

PC IVs 3.6.1.3 SURVE!1_L:ANCE ~EQUIRE~ENTS {continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.3 -------------------------------NOTES------------------------------

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

2. Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment manual isolation Prior to entering valve and blind flange that is located inside MODE 2 or 3 from primary containment and not locked, sealed, or MODE 4 if primary otherwise secured and is required to be closed containment was during accident conditions is closed. de-inerted while in MODE 4, if not performed within the previous 92 days

---*-*****--------- *-------*--*-**********--- ----------- - + - - - - - --*----*--

SR 3.6.1.3.4 Verify continuity of the traversing incore probe (TIP) In accordance with shear isolation valve explosive charge. the Surveillance Frequency Control Program SR 3.6.1.3.5 Verify the isolation time of each power operated, In accordance automatic PCIV, except for MS IVs, is within limits. with the INSERVICE TESTING PROGRAM (continued)

Cooper 3.6-13 Amendment No. 259

PC IVs 3.6.1.3

..§URVEJLL:ANCE Rs_QUIREiy1ENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify the isolation time of each MSIV is ~ 3 seconds In accordance and s 5 seconds. with the INSERVICE TESTING PROGRAM SR 3.6.1.3. 7 Verify each automatic PCIV actuates to the isolation In accordance with position on an actual or simulated isolation signal. the Surveillance Frequency Control Program SR 3.6.1.3.8 Verify a representative sample of reactor In accordance with instrumentation line EFCVs actuate to the Isolation the Surveillance position on an actual or simulated instrument line Frequency Control break. Program SR 3.6.1.3.9 Remove and test the explosive squib from each In accordance with shear isolation valve of the TIP System. the Surveillance Frequency Control Program

• -----*---*------------------* *---*~~ .

SR 3.6.1.3.10 Verify leakage rate through each Main Steam line is In accordance with s 106 scfh when tested at~ 29 psig. the Primary Containment Leakage Rate Testing Program (continued)

Cooper 3.6-14 Amendment No. 259

RHR Containment Spray 3.6.1.9 SURVEILLANCE REQUIREMENTS

--*~,~-*""'"" -***~'"'---=-

SURVEILLANCE FREQUENCY SR 3.6.1.9.1 Verify each RHR containment spray subsystem In accordance with manual, power operated, and automatic valve in the the Surveillance flow path that is not locked, sealed, or otherwise Frequency Control secured in position, is in the correct position or can be Program aligned to the correct position.

SR 3.6.1.9.2 Verify each required RHR pump develops a flow rate In accordance with of > 7700 gpm through the associated heat the INSERVICE exchanger while operating in the suppression pool TESTING cooling mode. PROGRAM SR 3.6.1.9.3 Verify each spray nozzle is unobstructed. Following maintenance which could result in nozzle blockage Cooper 3.6-26 Amendment No. 259

RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE I FREQUENCY

~*~* . **-1.

SR 3.6.2.3.1 Verify each RHR suppression pool cooling subsystem In accordance with manual, power operated, and automatic valve in the the Surveillance flow path that is not locked, sealed, or otherwise Frequency Control secured in position, is in the correct position or can Program be aligned to the correct position.

SR 3.6.2.3.2 Verify each RHR pump develops a flow rate > 7700 In accordance gpm through the associated heat exchanger while with the operating in the suppression pool cooling mode. INSERVICE TESTING PROGRAM Cooper 3.6-32 Amendment No. 259

SC IVs 3.6.4.2 SURVEILLANCE FREQUENCY SR 3.6.4.2.1 --NOTES--

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

2. Not required to be met for SCIVs that are open under administrative controls.

Verify each secondary containment isolation manual In accordance with valve and blind flange that is not locked, sealed, or the Surveillance otherwise secured and is required to be closed during Frequency Control accident conditions is closed. Program SR 3.6.4.2.2 Verify the isolation time of each power operated In accordance automatic SCIV is within limits. with the INSERVICE TESTING PROGRAM SR 3.6.4.2.3 Verify each automatic SCIV actuates to the isolation In accordance with position on an actual or simulated actuation signal. the Surveillance Frequency Control Program Cooper 3.6-39 Amendment No. 259

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.6 (Deleted)

(continued)

Cooper 5.0-10 Amendment No. 259

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 RELATED TO AMENDMENT NO. 259 TO FACILITY OPERATING LICENSE NO. DPR-46 NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298

1.0 INTRODUCTION

By application dated August 26, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16245A288), Nebraska Public Power District (NPPD, the licensee), requested changes to the technical specifications (TSs) for Cooper Nuclear Station (CNS). 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 CNS TS 5.5.6, "lnservice Testing Program," and add a new defined term, "INSERVICE TESTING PROGRAM," to the TSs. All existing references to the "lnservice Testing Program" in the CNS TS SRs are replaced with "INSERVICE TESTING PROGRAM" so that the SRs refer to the new definition in lieu of the deleted program.

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 American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (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 Enclosure 2

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, "Codes and standards," 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. Therefor~. 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 U.S. Nuclear Regulatory Commission (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 5.5.6 from the Administrative Controls section of TSs and replace it with the word "(Deleted)." TS 5.5.6 currently states:

This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 pumps and valves:

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

ASME OM Code and applicable Addenda Required Frequencies for terminology for inservice performing inservice testing testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every At least once per 92 days 3 months Semiannually or every At least once per 184 days 6 months Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every At least once per 731 days 2 years

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.

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. The licensee's proposed amendment is based on TSTF-545, Revision 3, which is an NRG-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 safety evaluation. 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 5.5.6 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 5.5.6.a The ASME OM Code requires testing to normally be performed within certain time periods.

TS 5.5.6.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. Therefore, the NRC staff determined that deletion of TS 5.5.6.a is acceptable.

Consideration of TS 5.5.6.b TS 5.5.6.b allows the licensee to extend, by up to 25 percent, the interval between inservice testing activities, as required by TS 5.5.6.a and for other normal and accelerated frequencies specified as 2 years or less in the inservice testing program. Similar to TS 5.5.6.b, the NRC authorization of ASME Code Case OMN-20, "lnservice Test Frequency," by letter dated February 12, 2016 (ADAMS Accession No. ML16014A174), 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 5.5.6.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 5.5.6.b is acceptable. The deletion of TS 5.5.6.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 5.5.6.c TS 5.5.6.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 5.5.6.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 5.5.6.c is acceptable.

Consideration of TS 5. 5. 6. d TS 5.5.6.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. 6 The NRC staff determined that the requirements currently in TS 5.5.6 are not necessary to assure operation of the facility in a safe manner. Based on this evaluation, the staff concludes that deletion of TS 5.5.6 from the licensee's TSs is acceptable, because TS 5.5.6 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 5.5.6.a. As discussed in Section 3.1 of this safety evaluation, the staff determined that the TSs do not need to include the more precise testing frequencies currently in TS 5.5.6.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:

• Since the CNS TS do not include SR 3.4.5.1, Reactor Coolant System Pressure Isolation Valve Leakage, as shown on the General Electric BWR [Boiling Water Reactor]/4 STS pages in TSTF-545, there will be no corresponding change to the CNS TS.

• General Electric BWR/4 STS SR 3.5.1.7 is numbered SR 3.5.1.6 in the CNS TS.

• General Electric BWR/4 STS SR 3.5.2.5 is numbered SR 3.5.2.4 in the CNS TS.

• General Electric BWR/4 STS SR 3.6.1.3.6 is numbered SR 3.6.1.3.5 in the CNS TS.

• General Electric BWR/4 STS SR 3.6.1.3.8 is numbered SR 3.6.1.3.6 in the CNS TS.

• General Electric BWR/4 STS section 3.6.2.4 is titled Residual Heat Removal (RHR) Suppression Pool Spray. The equivalent section in the CNS TS is numbered 3.6.1.9 and titled RHR Containment Spray.

• General Electric BWR/4 STS SR 3.6.2.4.2 is numbered SR 3.6.1.9.2 in the CNS TS.

• General Electric BWR/4 STS IST Program is section 5.5.7. The equivalent CNS TS section is 5.5.6.

• TSTF-545 deletes the IST [inservice testing] program TS 5.5.6 and re-numbers all subsequent TS programs. NPPD proposes to retain the TS 5.5.6 reference, now shown as "[(Deleted)]," and not change the subsequent TS program numbers.

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.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Nebraska State official was notified of the proposed issuance of the amendment on May 26, 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 November 8, 2016 (81 FR 78649). 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: June 20, 2017

ML17144A082 *via e-mail OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DORL/LPL4/PM NRR/DSS/STSB/BC(A)*

NAME APulvirenti PBlechman TWengert JWhitman DATE 5/26/17 5/25/17 5/30/17 6/5/17 OFFICE NRR/DE/EPNB/BC OGG (NLO) NRR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME DAiiey AGhosh RPascarelli TWengert DATE 5/31/17 6/15/17 6/19/17 6/20/17