ML17123A321

Issuance of Amendment Adoption of TSTF-545, Revision 3, TS Inservice Testing Program Removal and Clarify SR Usage Rule Application to Section 5.5 Testing

ML17123A321
Person / Time
Site:	Monticello
Issue date:	06/16/2017
From:	Robert Kuntz Plant Licensing Branch III
To:	Gardner P Northern States Power Company, Minnesota
Kuntz R, DORL/LPLIII, 415-3733
References
CAC MF8217
Download: ML17123A321 (25)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 16, 2017 Mr. Peter A Gardner Site Vice President Monticello Nuclear Generating Plant Northern States Power Company - Minnesota (NSPM) 2807 West County Road 75 Monticello, MN 55362-9637

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT - ISSUANCE OF AMENDMENT RE: ADOPTION OF TSTF-545, REVISION 3, "TS INSERVICE TESTING PROGRAM REMOVAL & CLARIFY SR USAGE RULE APPLICATION TO SECTION 5.5 TESTING" (CAC NO. MF8217)

Dear Mr. Gardner:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 194 to Renewed Facility Operating License No. DPR-22 for the Monticello Nuclear Generating Plant.

The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated July 28, 2016.

The amendment adopts TSTF-545, Revision 3, "TS lnservice Testing Program Removal &

Clarify SR [surveillance requirements] Usage Rule Application to Section 5.5 Testing."

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

Sincerely,

~.

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R tz, Senior Project Manager Plant Licensing Branch 111 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-263

Enclosures:

1. Amendment No. 194 to DPR-22

2. Safety Evaluation cc: ListServ

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 194 License No. DPR-22

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

A. The application for amendment by Northern States Power Company (NSPM, the licensee), dated July 28, 2016 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes 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-22 is hereby amended to read as follows:

2. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 194, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULA TORY COMMISSION

()J ~V-David J. Wrona, Chief Plant Licensing Branch 111 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Operating License No. DPR-22 and Technical Specifications Date of Issuance: June 16, 2017

ATTACHMENT TO LICENSE AMENDMENT NO. 194 MONTICELLO NUCLEAR GENERATING PLANT RENEWED FACILITY OPERATING LICENSE NO. DPR-22 DOCKET NO. 50-263 Replace the following page of Renewed Facility Operating License DPR-22 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

REMOVE INSERT 3 3 Replace the following pages of 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.

REMOVE INSERT 1.1-2 1.1-2 1.1-3 1.1-3 3.1.7-3 3.1.7-3 3.4.3-2 3.4.3-2 3.5.1-5 3.5.1-5 3.5.1-6 3.5.1-6 3.5.1-7 3.5.1-7 3.5.2-3 3.5.2-3 3.5.3-2 3.5.3-2 3.6.1.5-2 3.6.1.5-2 3.6.2.3-2 3.6.2.3-2 5.5-4 5.5-4

3

2. Pursuant to the Act and 10 CFR Part 70, NSPM to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operations, as described in the Final Safety Analysis Report, as supplemented and amended, and the licensee's filings dated August 16, 1974 (those portions dealing with handling of reactor fuel);

3. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM 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;

4. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM 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

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

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations in 10 CFR Chapter I 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 NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 2004 megawatts (thermal).

2. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 194, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications.

3. Physical Protection NSPM shall implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Renewed License No. DPR-22 Amendment No. 194

Definitions 1.1 1.1 Definitions CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

The following exceptions are not considered to be CORE ALTERATIONS:

a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement); and

b. Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that provides cycle REPORT (COLR) specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.3. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The dose conversion factors used for this calculation shall be those listed in Federal Guidance Report (FGR)-11, "Limiting Values of Radionuclide Intake and Air Concentration Factors for Inhalation, Submersion and Ingestion," September 1988, and FGR-12, "External Exposure to Radionuclides in Air, Water and Soil,"

September 1993.

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 drywell, such as that from pump seals or valve packing that is captured and conducted to a sump or collecting tank; or Monticello 1.1-2 Amendment No. 146, 148, 194 I

Definitions 1.1 1.1 Definitions LEAKAGE (continued)

2. LEAKAGE into the drywell 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; and

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.

LOGIC SYSTEM A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all FUNCTIONAL 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) that RATIO (MCPR) 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.

Monticello 1.1-3 Amendment No. 44e, 194 I

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.7 Verify each pump develops a flow rate ;;::: 24 gpm at In accordance a discharge pressure ;;::: 1275 psig. with the INSERVICE TESTING PROGRAM SR 3.1.7.8 Verify flow through one SLC subsystem from pump 24 months on a into reactor pressure vessel. STAGGERED TEST BASIS SR 3.1.7.9 Verify all heat traced piping between storage tank 24 months and pump suction is unblocked.

AND

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NOTE-------

Only required if SLC pump suction lines heat tracing is inoperable.

Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after room temperature in the vicinity of the SLC pumps is restored within the solution temperature limits of Figure 3.1. 7-2 SR 3.1.7.10 Verify sodium pentaborate enrichment is Prior to addition

55.0 atom percent 8-10. to SLC tank Monticello 3.1.7-3 Amendment No. 449, 194 I

S/RVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints of the required In accordance S/RVs are 1109 +/-. 33.2 psig. Following testing, lift with the settings shall be 1109 +/-. 11.0 psig. INSERVICE TESTING PROGRAM 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 flow is adequate to perform the test.

Verify each required S/RV is capable of being In accordance opened. with the INSERVICE TESTING PROGRAM Monticello 3.4.3-2 Amendment No. 146, 168, 194 I

ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify, for each ECCS injection/spray subsystem, 31 days locations susceptible to gas accumulation are sufficiently filled with water.

SR 3.5.1.2 -------------------------------NOTE------------------------------

Not required to be met for system vent flow paths opened under administrative control.

Verify each ECCS injection/spray subsystem 31 days manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.1.3 Verify ADS pneumatic pressure is as follows for 31 days each required ADS pneumatic supply:

a. S/RV Accumulator Bank header pressure

~ 88.3 psig; and

b. Alternate Nitrogen System pressure is

~ 1060 psig.

SR 3.5.1.4 -------------------------------NOTE------------------------------

Only required to be met in MODE 1.

Verify the RHR System intertie return line isolation 31 days valves are closed.

SR 3.5.1.5 Verify correct breaker alignment to the LPCI swing 31 days bus.

SR 3.5.1.6 Verify each recirculation pump discharge valve In accordance cycles through one complete cycle of full travel or is with the de-energized in the closed position. INSERVICE TESTING PROGRAM Monticello 3.5.1-5 Amendment No. 146, 155, 162, 189, 190, 194

ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.7 Verify the following ECCS pumps develop the In accordance specified flow rate against a system head with the corresponding to the specified reactor to INSERVICE containment pressure. TESTING PROGRAM System Head Corresponding to a Reactor to No. of Containment System Flow Rate Pumps Pressure of Core Spray ~ 2835 gpm 1 ~ 130 psi LPCI ~ 3870 gpm 1 ~ 20 psi SR 3.5.1.8 -------------------------------N()TE------------------------------

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 steam dome pressure In accordance

~ 1025.3 psig and ~ 950 psig, the HPCI pump can with the develop a flow rate ~ 2700 gpm against a system INSERVICE head corresponding to reactor pressure. TESTING PROGRAM SR 3.5.1.9 -------------------------------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~ 165 psig, the HPCI 24 months pump can develop a flow rate ~ 2700 gpm against a system head corresponding to reactor pressure.

Monticello 3.5.1-6 Amendment No. 146, 162, 167, 194 I

ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.10 -------------------------------NOTE------------------------------

Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem 24 months actuates on an actual or simulated automatic initiation signal.

SR 3.5.1.11 -------------------------------N()TE------------------------------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or simulated 24 months automatic initiation signal.

SR 3.5.1.12 -------------------------------N()TE------------------------------

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 flow is adequate to perform the test.

Verify each ADS valve is capable of being opened. In accordance with the INSERVICE TESTING PROGRAM SR 3.5.1.13 Verify automatic transfer capability of the LPCI 24 months swing bus power supply from the normal source to the backup source.

Monticello 3.5.1-7 AmendmentNo.146, 162, 168, 194 I

ECCS - Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.2 Verify, for each required ECCS injection/spray 31 days subsystem, locations susceptible to gas accumulation are sufficiently filled with water.

SR 3.5.2.3 -------------------------------N()TE------------------------------

Not required to be met for system vent flow paths opened under administrative control.

Verify each required ECCS injection/spray 31 days subsystem manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.2.4 Verify each required ECCS pump develops the In accordance specified flow rate against a system head with the corresponding to the specified reactor to INSERVICE containment pressure. TESTING PR()GRAM System Head Corresponding to a Reactor to No. of Containment System Flow Rate Pumps Pressure of Core Spray ~ 2835 gpm 1 ~ 130 psi LPCI ~ 3870 gpm 1 ~ 20 psi SR 3.5.2.5 -------------------------------N()TE------------------------------

Vessel injection/spray may be excluded.

Verify each required ECCS injection/spray 24 months subsystem actuates on an actual or simulated automatic initiation signal.

Monticello 3.5.2-3 Amendment No. 146, 167, 189, 194 I

RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 -------------------------------N()TE------------------------------

Not required to be met for system vent flow paths opened under administrative control.

Verify each RCIC System manual, power operated, 31 days and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.3.2 -------------------------------N()TE------------------------------

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 1025.3 psig and In accordance

~ 950 psig, the RCIC pump can develop a flow rate with the

~ 400 gpm against a system head corresponding to INSERVICE reactor pressure. TESTING PR()GRAM SR 3.5.3.3 -------------------------------N()TE------------------------------

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 165 psig, the RCIC 24 months pump can develop a flow rate ~ 400 gpm against a system head corresponding to reactor pressure.

SR 3.5.3.4 -------------------------------N()TE------------------------------

Vessel injection may be excluded.

Verify the RCIC System actuates on an actual or 24 months simulated automatic initiation signal.

SR 3.5.3.5 Verify the RCIC System locations susceptible to gas 31 days accumulation are sufficiently filled with water.

Monticello 3.5.3-2 Amendment No. 146, 189, 194 I

LLS Valves 3.6.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.5.1 -------------------------------N()TE------------------------------

N ot 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 flow is adequate to perform the test.

Verify each LLS valve is capable of being opened. In accordance with the INSERVICE TESTING PR()GRAM SR 3.6.1.5.2 -------------------------------N()TE------------------------------

Valve actuation may be excluded.

Verify the LLS System actuates on an actual or 24 months simulated automatic initiation signal.

Monticello 3.6.1.5-2 Amendment No. 146, 168, 194 I

RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.2.3.2 Verify each required RHR pump develops a flow In accordance rate ~ 3870 gpm through the associated heat with the exchanger while operating in the suppression pool INSERVICE cooling mode. TESTING PROGRAM SR 3.6.3.2.3 Verify RHR suppression pool cooling subsystem 31 days locations susceptible to gas accumulation are sufficiently filled with water Monticello 3.6.2.3-2 Amendment No. 146, 189, 194 I

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.5 (Deleted) 5.5.6 Ventilation Filter Testing Program (VFTP)

A program shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems. Tests described in Specifications 5.5.6.a and 5.5.6.b shall be performed once per 24 months and following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the high efficiency particulate air (HEPA) filters or charcoal adsorber capability.

Monticello 5.5-4 Amendment No. -Me, 194 I

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 194 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-22 NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263

1.0 INTRODUCTION

By application dated July 28, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16210A030), Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (the licensee), requested changes to the technical specifications (TSs) for Monticello Nuclear Generating Plant (MNGP). 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 MNGP TS 5.5.5, "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 MNGP 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's letter dated July 28, 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 MNGP. The U.S. Nuclear Regulatory Commission (NRC or Commission) considered this request separately from the proposed license amendment, and authorized the licensee's use of this alternative by letter dated May 15, 2017 (ADAMS Accession No. ML17122A157).

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 Enclosure 2

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, "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. 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 5.5.5 from the Administrative Controls section of TSs and replace it with "(Deleted)." TS 5.5.5 states that the program provides controls for inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. TS 5.5.5.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").

TS 5.5.5.b states that the provisions of SR 3.0.2 apply to the frequencies in TS 5.5.5.a.

SR 3.0.2 allows an extension of inservice testing intervals by up to 25 percent. TS 5.5.5.c states that the provisions of SR 3.0.3 are applicable to inservice testing activities. 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. TS 5.5.5.d states that nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

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.

/nservice 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 5.5.5 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. 5. a The ASME OM Code requires testing to normally be performed within certain time periods.

TS 5.5.5.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. 5. b TS 5.5.5.b allows the licensee to extend, by up to 25 percent, the interval between inservice testing activities, as required by TS 5.5.5.a. Similar to TS 5.5.5.b, the NRC authorization of ASME Code Case OMN-20, "lnservice Test Frequency," by letter dated May 15, 2017, 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 5.5.5.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 5.5.5.b is acceptable. The deletion of TS 5.5.5.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. 5. c TS 5.5.5.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 5.5.5.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 5.5.5.c is acceptable.

Consideration of TS 5. 5. 5. d TS 5.5.5.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. 5 The NRC staff determined that the requirements currently in TS 5.5.5 are not necessary to assure operation of the facility in a safe manner. Based on this evaluation, the staff concludes that deletion of TS 5.5.5 from the licensee's TSs is acceptable, because TS 5.5.5 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 5.5.5.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 5.5.5.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 5.5.5 from the TSs and renumbers the subsequent TS programs. The licensee proposes to delete the content of TS 5.5.5, but retains the TS number, and adds "(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.

3. In the letter, the licensee stated the following:

The MNGP TS do not include Specification 3.4.5, "RCS [Reactor Coolant System] Primary Isolation Valve Leakage" that is contained in NUREG-1433. Additionally, the MNGP TS do not include Specification 3.6.2.4, "Residual Heat Removal (RHR)

Suppression Pool Spray" that is contained in NUREG-1433. The MNGP TS do include another RHR specification for a similar application, Specification 3.6.1.8, "RHR Drywell Spray". However, MNGP TS 3.6.1.8 does not include a commensurate SR to 3.6.2.4.2, and therefore a commensurate change is not required.

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 Minnesota State official was notified of the proposed issuance of the amendment on May 3, 2017. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

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 October 11, 2016 (81 FR 70181 ). 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: Caroline Tilton, NRR Date of issuance: June 16, 2017

ML17123A321 *via e-mail OFFICE DORL/LPL3/PM DORL/LPL3/LA DSS/STSB/BC(A)* DSS/SNPB/BC*

NAME RKuntz SRohrer JWhitman DAiiey DATE 05/4/17 05/4/17 05/11/17 05/11/17 OFFICE OGC(NLO) DORL/LPL3/BC DORL/LPL3/PM NAME BMizuno DWrona RKuntz DATE 05/31/17 06/16/17 06/16/17