ML17103A106
| ML17103A106 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/24/2017 |
| From: | Jennivine Rankin Plant Licensing Branch III |
| To: | Gebbie J P Indiana Michigan Power Co |
| Dietrich A P | |
| References | |
| CAC MF8156, CAC MF8157 | |
| Download: ML17103A106 (40) | |
Text
Mr. Joel P. Gebbie Senior Vice President and Chief Nuclear Officer UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 24, 2017 Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, Ml 49106
SUBJECT:
DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 -ISSUANCE OF AMENDMENTS RE: ADOPTION OF TSTF-545, REVISION 3, "TS INSERVICE TESTING PROGRAM REMOVAL & CLARIFY SR USAGE RULE APPLICATION TO SECTION 5.5 TESTING" (CAC NOS. MF8156 AND MF8157)
Dear Mr. Gebbie:
The U.S. Nuclear Regulatory Commission (NRC or Commission) has issued the enclosed Amendment No. 335 to Renewed Facility Operating License No. DPR-58 and Amendment No. 317 to Renewed Facility Operating License No. DPR-74 for the Donald C. Cook Nuclear Plant (CNP), Units 1 and 2, respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated July 21, 2016, as supplemented by letter dated September 26, 2016. The amendments revise technical specifications (TSs) surveillance requirements (SRs), consistent with the NRG-approved Technical Specifications Task Force (TSTF) Traveler, TSTF-545, Revision 3, "TS lnservice Testing [IST] Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing." Specifically, the change revises the TSs to eliminate Section 5.5.6, "lnservice Testing Program." A new defined term, "INSERVICE TESTING PROGRAM," is added to the TS Definitions Section. The TS SRs that currently refer to the "lnservice Testing Program" from Section 5.5.6 are revised to refer to the new defined term, "INSERVICE TESTING PROGRAM."
J. Gebbie A copy of our related safety evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Docket Nos. 50-315 and 50-316
Enclosures:
- 1. Amendment No. 335 to DPR-58 2. Amendment No. 317 to DPR-74 3. Safety Evaluation cc w/encls: Distribution via ListServ Sincerely, nivine Rankin, Project Manager Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-315 DONALD C. COOK NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 335 License No. DPR-58 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Indiana Michigan Power Company (the licensee) dated July 21, 2016, as supplemented by letter dated September 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 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-58 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 335, are hereby incorporated in this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days.
Attachment:
Changes to Renewed Facility Operating License No. DPR-58 and Technical Specifications Date of Issuance: May 24, 201 7 FOR THE NUCLEAR REGULATORY COMMISSION OJ 9* r ____ . David J. Wrona, Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ATTACHMENT TO LICENSE AMENDMENT NO. 335 DONALD C. COOK NUCLEAR PLANT. UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-58 DOCKET NO. 50-315 Replace the following page of the Renewed Facility Operating License No. DPR-58 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 1.1-3 3.4.10-2 3.4.14-2 3.5.2-2 3.6.3-5 3.6.6-1 3.7.1-2 3.7.2-2 3.7.3-2 3.7.5-3 5.5-4 INSERT 1.1-3 3.4.10-2 3.4.14-2 3.5.2-2 3.6.3-5 3.6.6-1 3.7.1-2 3.7.2-2 3.7.3-2 3.7.5-3 5.5-4 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, 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 and equipment calibration or associated with radioactive apparatus or components; and (5) 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 operating 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; 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 The licensee is authorized to operate the facility at steady state reactor core power levels not to exceed 3304 megawatts thermal in accordance with the conditions specified herein. (2) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 335, are hereby incorporated in this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. (3) Less than Four Loop Operation The licensee shall not operate the reactor at power levels above P-7 (as defined in Table 3.3.1-1 of Specification 3.3.1 of Appendix A to this renewed operating license) with less than four reactor coolant loops in operation until (a) safety analyses for less than four loop operation have been submitted, and (b) approval for less than four loop operation at power levels above P-7 has been granted by the Commission by amendment of this license. (4) Fire Protection Program Indiana Michigan Power Company 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 licensee's amendment request dated July 1, 2011, as supplemented by letters dated September 2, 2011, April 27, 2012, June 29, 2012, August 9, 2012, October 15, 2012, November9, 2012, January 14, 2013, February 1, 2013, Renewed License No. DPR-58 Amendment JU, J.2g, JJ4, 335 1.1 Definitions ENGINEERED SAFETY FEATURE(ESF)RESPONSE TIME INSERVICE TESTING PROGRAM LEAKAGE Cook Nuclear Plant Unit 1 Definitions 1.1 The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. 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. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC. The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 1 O CFR 50.55a(f). LEAKAGE shall be: a. Identified LEAKAGE 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (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 or leakoff) that is not identified LEAKAGE; and c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall. 1.1-3 Amendment No. 28-7, 200, 335 Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SR 3.4.10.1 SURVEILLANCE Verify each pressurizer safety valve is OPERABLE in accordance with the INSERVICE TESTING PROGRAM. Following testing, lift settings shall be within+/- 1%. FREQUENCY In accordance with the INSERVICE TESTING PROGRAM Cook Nuclear Plant Unit 1 3.4.10-2 Amendment No. 335 RCS PIV Leakage 3.4.14 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME A.2 Isolate the high pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> portion of the affected system from the low pressure portion by use of a second closed manual, deactivated automatic, or check valve. 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 of Condition A not AND met. B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. RHR System interlock C.1 Isolate the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> function inoperable. penetration by use of one closed manual or deactivated automatic valve. SURVEILLANCE REQUIREMENTS SR 3.4.14.1 SURVEILLANCE -----------------------NOTE---------------------------On ly required to be performed in MODES 1 and 2. Verify leakage from each RCS PIV is equivalent to s: 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure :::: 2215 psig and s: 2255 psig. FREQUENCY In accordance with the INSERVICE TESTING PROGRAM Cook Nuclear Plant Unit 1 3.4.14-2 Amendment No. 335 SURVEILLANCE REQUIREMENTS SR 3.5.2.1 SR 3.5.2.2 SR 3.5.2.3 SR 3.5.2.4 SR 3.5.2.5 SURVEILLANCE Verify the following valves are in the listed position with power to the valve operator removed. Number 1-IM0-261 1-IM0-262 1-IM0-263 1-IM0-315 1-IM0-325 1-IM0-390 1-ICM-305 1-ICM-306 Position Open Open Open Closed Closed Open Closed Closed Function SI suction line Mini flow line Mini flow line Low head SI to hot leg Low head SI to hot leg RWSTto RHR Sump line Sump line ------------------------------N 0 TE----------------------------Not required to be met for system vent flow paths opened under administrative control. Verify each ECCS 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. Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head. Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. Verify each ECCS pump starts automatically on an actual or simulated actuation signal. ECCS -Operating 3.5.2 FREQUENCY In accordance with the Surveillance Frequency Control Program 1n accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Cook Nuclear Plant Unit 1 3.5.2-2 Amendment No. -dM, 335 Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued) SR 3.6.3.4 SR 3.6.3.5 SURVEILLANCE Verify the isolation time of each automatic power operated containment isolation valve is within limits. Verify each automatic containment isolation valve that is not locked, sealed, or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal. FREQUENCY In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Cook Nuclear Plant Unit 1 3.6.3-5 Amendment No. 2S7, 335 3.6 CONT Al NM ENT SYSTEMS 3.6.6 Containment Spray System Containment Spray System 3.6.6 LCO 3.6.6 Two containment spray trains shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION A. One containment spray A.1 Restore containment spray train inoperable. train to OPERABLE status. B. Required Action and 8.1 Be in MODE 3. associated Completion Time not met. AND B.2 Be in MODE 5. SURVEILLANCE REQUIREMENTS SR 3.6.6.1 SURVEILLANCE -----------------------------N 0 TE---------------------------Not required to be met for system vent flow paths opened under administrative control. Verify each containment spray 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. COMPLETION TIME 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 6 hours 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> FREQUENCY In accordance with the Surveillance Frequency Control Prooram SR 3.6.6.2 Verify each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head. In accordance with the INSERVICE TESTING PROGRAM Cook Nuclear Plant Unit 1 3.6.6-1 Amendment No. 281, 334, 335 ACTIONS (continued) CONDITION REQUIRED ACTION B. Required Action and B.1 Be in MODE 3. associated Completion Time of Condition A not AND met. One or more steam generators with 4 MSSVs inoperable. B.2 SURVEILLANCE REQUIREMENTS Be in MODE4. SURVEILLANCE MSSVs 3.7.1 COMPLETION TIME 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours FREQUENCY SR 3.7.1.1 ------------------------------NOTE------------------------------On ly required to be performed in MODES 1 and 2. Verify each required MSSV is OPERABLE in accordance with the INSERVICE TESTING PROGRAM with the lift setpoint per Table 3.7.1-2. Following testing, lift setting shall be within +/-1 %. In accordance with the INSERVICE TESTING PROGRAM Cook Nuclear Plant Unit 1 3.7.1-2 Amendment No. 2&7, 335 SURVEILLANCE REQUIREMENTS SR 3.7.2.1 SR 3.7.2.2 SURVEILLANCE ------------------------------N 0 TE------------------------------On ly required to be performed in MODES 1 and 2. Verify the isolation time of each SGSV is within limits. -------------------------------NOTE------------------------------Only required to be performed in MODES 1 and 2. Verify each SGSV actuates to the isolation position on an actual or simulated actuation signal. SGSVs 3.7.2 FREQUENCY In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Cook Nuclear Plant Unit 1 3.7.2-2 Amendment 335 ACTIONS {continued) CONDITION REQUIRED ACTION MFIVs and MFRVs 3.7.3 COMPLETION TIME D. Required Action and associated Completion Time not met. D.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 /> AND D.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 SR 3.7.3.1 SR 3.7.3.2 SR 3.7.3.3 SURVEILLANCE Verify the isolation time of each MFIV is within limits. Verify the isolation time of each MFRV is within limits. Verify each MFIV and MFRV actuates to the isolation position on an actual or simulated actuation signal. FREQUENCY In accordance with the INSERVICE TESTING PROGRAM In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Cook Nuclear Plant Unit 1 3.7.3-2 Amendment No. ;IB7, J.4-8, JM, 335 SURVEILLANCE REQUIREMENTS SR 3.7.5.1 SR 3.7.5.2 SR 3.7.5.3 SURVEILLANCE -------------------------------NO TE-----------------------------AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation. Verify each required AFW manual, power operated, and 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. -------------------------------NO TE-----------------------------N ot required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after<:: 850 psig in the steam generator. Verify the developed head of each required AFW pump at the flow test point is greater than or equal to the required developed head. -------------------------------N 0 TES----------------------------1. AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation. 2. Only required to be met in MODES 1, 2, and 3. Verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. AFW System 3.7.5 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Cook Nuclear Plant Unit 1 3.7.5-3 Amendment No. 335 5.5 Programs and Manuals 5.5.5 Reactor Coolant Pump Flywheel Inspection Program Programs and Manuals 5.5 This program shall provide for the inspection of each reactor coolant pump flywheel. A qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (magnetic particle testing or penetrant testing, or combination of the two tests) of exposed surfaces of the removed flywheels shall be conducted once every 10 years. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program Surveillance Frequency. 5.5.6 DELETED Cook Nuclear Plant Unit 1 5.5-4 Amendment No. 335 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-316 DONALD C. COOK NUCLEAR PLANT. UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 317 License No. DPR-74 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Indiana Michigan Power Company (the licensee) dated July 21, 2016, as supplemented by letter dated September 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 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 2 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-74 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 317, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days.
Attachment:
Changes to Renewed Facility Operating License No. DPR-74 and Technical Specifications Date of Issuance: May 24, 201 7 FOR THE NUCLEAR REGULATORY COMMISSION C)J 9 4/ ___ David J. Wrona, Chief Plant Licensing Branch 111 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ATTACHMENT TO LICENSE AMENDMENT NO. 317 DONALD C. COOK NUCLEAR PLANT. UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-74 DOCKET NO. 50-316 Replace the following page of the Renewed Facility Operating License No. DPR-74 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 1.1-3 3.4.10-2 3.4.14-2 3.5.2-2 3.6.3-5 3.6.6-1 . 3.7.1-2 3.7.2-2 3.7.3-2 3.7.5-3 5.5-4 INSERT 1.1-3 3.4.10-2 3.4.14-2 3.5.2-2 3.6.3-5 3.6.6-1 3.7.1-2 3.7.2-2 3.7.3-2 3.7.5-3 5.5-4 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, 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 and equipment calibration or associated with radioactive apparatus or components; and (5) 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 faciltty. C. This renewed operating 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; 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 The licensee is authorized to operate the facility at steady state reactor core power levels not to exceed 3468 megawatts thermal in accordance with the conditions specified herein and in Attachment 1 to the renewed operating license. The preoperational tests, startup tests and other items identified in Attachment 1 to this renewed operating license shall be completed. Attachment 1 is an integral part of this renewed operating license. (2) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 317, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. (3) Additional Conditions (a) Deleted by Amendment No. 76 (b) Deleted by Amendment No. 2 (c) Leak Testing of Emergency Core Cooling System Valves Indiana Michigan Power Company shall prior to completion of the first inservice testing interval leak test each of the two valves in series in the Renewed License No. DPR-74 Amendment No., WQ, J..:t.-0, J..:t.2, 344, 317 1.1 Definitions ENGINEERED SAFETY FEATURE(ESF)RESPONSE TIME INSERVICE TESTING PROGRAM LEAKAGE Cook Nuclear Plant Unit 2 Definitions 1.1 The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. 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. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC. The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f). LEAKAGE shall be: a. Identified LEAKAGE 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (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 or leakoff) that is not identified LEAKAGE; and c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall. 1.1-3 Amendment No. 2W, 317 Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SR 3.4.10.1 SURVEILLANCE Verify each pressurizer safety valve is OPERABLE in accordance with the INSERVICE TESTING PROGRAM. Following testing, lift settings shall be within+/- 1%. FREQUENCY In accordance with the INSERVICE TESTING PROGRAM Cook Nuclear Plant Unit 2 3.4.10-2 Amendment 317 RCS PIV Leakage 3.4.14 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME A.2 Isolate the high pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> portion of the affected system from the low pressure portion by use of a second closed manual, deactivated automatic, or check valve. 8. 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 of Condition A not AND met. 8.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. RHR System interlock C.1 Isolate the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> function inoperable. penetration by use of one closed manual or deactivated automatic valve. SURVEILLANCE REQUIREMENTS SR 3.4.14.1 SURVEILLANCE -----------------------------N 0 TE----------------------------Only required to be performed in MODES 1 and 2. Verify leakage from each RCS PIV is equivalent to s 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS 2215 psig and :S 2255 psig. FREQUENCY In accordance with the INSERVICE TESTING PROGRAM Cook Nuclear Plant Unit 2 3.4.14-2 Amendment 317 SURVEILLANCE REQUIREMENTS SR 3.5.2.1 SR 3.5.2.2 SR 3.5.2.3 SR 3.5.2.4 SR 3.5.2.5 SURVEILLANCE Verify the following valves are in the listed position with power to the valve operator removed. Number 2-IM0-261 2-IM0-262 2-IM0-263 2-IM0-315 2-IM0-325 2-IM0-390 2-ICM-305 2-ICM-306 Position Open Open Open Closed Closed Open Closed Closed Function SI suction line Mini flow line Mini flow line Low head SI to hot leg Low head SI to hot leg RWSTto RHR Sump line Sump line ----------------------------NOTE--------------------------Not required to be met for system vent flow paths opened under administrative control. Verify each ECCS 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. Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head. Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. Verify each ECCS pump starts automatically on an actual or simulated actuation signal. ECCS -Operating 3.5.2 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Cook Nuclear Plant Unit 2 3.5.2-2 Amendment No. 200, J.+e, 317 Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued) SR 3.6.3.4 SR 3.6.3.5 SURVEILLANCE Verify the isolation time of each automatic power operated containment isolation valve is within limits. Verify each automatic containment isolation valve that is not locked, sealed, or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal. FREQUENCY In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Cook Nuclear Plant Unit 2 3.6.3-5 Amendment 317 Containment Spray System 3.6.6 3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Spray System LCO 3.6.6 Two containment spray trains shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION A. One containment spray A.1 Restore containment spray train inoperable. train to OPERABLE status. B. Required Action and B.1 Be in MODE 3. associated Completion Time not met. AND B.2 Be in MODE 5. SURVEILLANCE REQUIREMENTS SR 3.6.6.1 SURVEILLANCE Not required to be met for system vent flow paths opened under administrative control. Verify each containment spray 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. COMPLETION TIME 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 6 hours 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> FREQUENCY In accordance with the Surveillance Frequency Control Program SR 3.6.6.2 Verify each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head. In accordance with the INSERVICE TESTING PROGRAM Cook Nuclear Plant Unit 2 3.6.6-1 Amendment 317 ACTIONS (continued) CONDITION REQUIRED ACTION B. Required Action and B.1 Be in MODE 3. associated Completion Time of Condition A not AND met. One or more steam generators with ;:: 4 MSSVs inoperable. B.2 SURVEILLANCE REQUIREMENTS Be in MODE 4. SURVEILLANCE MSSVs 3.7.1 COMPLETION TIME 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours FREQUENCY SR 3.7.1.1 ------------------------------N 0 TE------------------------------Only required to be performed in MODES 1 and 2. Verify each required MSSV is OPERABLE in accordance with the INSERVICE TESTING PROGRAM with the lift setpoint per Table 3.7.1-2. Following testing, lift setting shall be within +/-1%. In accordance with the INSERVICE TESTING PROGRAM Cook Nuclear Plant Unit 2 3.7.1-2 Amendment 317 SURVEILLANCE REQUIREMENTS SR 3.7.2.1 SR 3.7.2.2 SURVEILLANCE ------------------------------NO TE----------------------------On ly required to be performed in MODES 1 and 2. Verify the isolation time of each SGSV is within limits. --------------------------NOTE-----------------------------On ly required to be performed in MODES 1 and 2. Verify each SGSV actuates to the isolation position on an actual or simulated actuation signal. SGS Vs 3.7.2 FREQUENCY In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Cook Nuclear Plant Unit 2 3.7.2-2 Amendment No. 269, JG-1-, J-1..e, 317 ACTIONS (continued) CONDITION REQUIRED ACTION MFIVs and MFRVs 3.7.3 COMPLETION TIME D. Required Action and associated Completion Time not met. D.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 /> D.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 SR 3.7.3.1 SR 3.7.3.2 SR 3.7.3.3 SURVEILLANCE Verify the isolation time of each MFIV is within limits. Verify the isolation time of each MFRV is within limits. Verify each MFIV and MFRV actuates to the isolation position on an actual or simulated actuation signal. FREQUENCY In accordance with the INSERVICE TESTING PROGRAM In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Cook Nuclear Plant Unit 2 3.7.3-2 Amendment 317 SURVEILLANCE REQUIREMENTS SR 3.7.5.1 SR 3.7.5.2 SR 3.7.5.3 SURVEILLANCE ----------------------------NO TE----------------------------AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation. Verify each required AFW manual, power operated, and 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. ------------------------------NO TE------------------------------Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after 2: 850 psig in the steam generator. Verify the developed head of each required AFW pump at the flow test point is greater than or equal to the required developed head. ------------------------------N 0 TES---------------------------1. AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation. 2. Only required to be met in MODES 1, 2, and 3. Verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. AFW System 3.7.5 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Cook Nuclear Plant Unit 2 3.7.5-3 Amendment 317 5.5 Programs and Manuals 5.5.5 Reactor Coolant Pump Flywheel Inspection Program Programs and Manuals 5.5 This program shall provide for the inspection of each reactor coolant pump flywheel. A qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (magnetic particle testing or penetrant testing, or combination of the two tests) of exposed surfaces of the removed flywheels shall be conducted once every 10 years. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program Surveillance Frequency. 5.5.6 DELETED Cook Nuclear Plant Unit 2 5.5-4 Amendment No. 299, 317 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 335 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 AND AMENDMENT NO. 317 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-315 AND 50-316
1.0 INTRODUCTION
By application dated July 21, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 16208A076), as supplemented by letter dated September 26, 2016 (ADAMS Accession No. ML 16272A 165), Indiana Michigan Power Company, LLC (l&M, the licensee) requested license amendments for the Donald C. Cook Nuclear Plant (CNP), Unit Nos. 1 and 2. The proposed amendments would revise technical specifications (TSs) consistent with Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-545, Revision 3, "TS lnservice Testing [IST] Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing," dated October 21, 2015 (ADAMS Accession No. ML 15294A555). The licensee's proposed changes delete CNP 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 CNP TS SRs are replaced with "INSERVICE TESTING PROGRAM" so that the SRs refer to the new definition in lieu of the deleted program. The supplemental letter dated September 26, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register(FR) on September 27, 2016 (81 FR 66307). The letter dated July 21, 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 CNP. 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 April 12, 2017 (ADAMS Accession No. ML 17096A627). Enclosure 3
2.0 REGULATORY EVALUATION
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 IST 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 IST 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 IST 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. ML 15317A071), and published a notice of availability in the 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 specified in the ASME Operation and Maintenance Standards and Guides (OM Codes) and applicable Addenda are as follows:
ASME OM Codes and applicable Addenda terminology for inservice testing activities Quarterly or every 3 months Yearly or annually Biennially or every 2 years Required Frequencies for performing inservice testing activities At least once per 92 days At least once per 366 days At least once per 731 days b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities. c. The provisions of SR 3.0.3 are applicable to inservice testing activities. d. Nothing in the ASME OM Codes shall be construed to supersede the requirements of any TS. TS 5.5.6.b, which references SR 3.0.2, allows an extension of IST intervals by up to 25 percent. If it is discovered that a surveillance associated with an IST 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. ML 100351425). 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 1 O CFR Part 50. These requirements include IST 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 1 O 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 IST 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. ML 13295A020) provides guidance for the IST 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 IST 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 IST program that provides controls for IST of ASME Code Class 1, 2, and 3 components (i.e., pumps and valves). Through 10 CFR 50.54, the applicable requirements of 1 O CFR 50.55a are conditions of every nuclear power reactor operating license issued under 1 O CFR Part 50. These requirements include 1 O CFR 50.55a(f), which specifies the requirements for the IST of pumps and valves. Therefore, requiring the licensee to have an IST 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 IST 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 IST 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 IST 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 IST frequencies are not necessary to assure operation of the facility in a safe manner. 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 IST activities, as required by TS 5.5.6.a. Similar to TS 5.5.6.b, the NRC authorization of ASME Code Case OMN-20, "lnservice Test Frequency," by letter dated April 12, 2017 (ADAMS Accession No. ML 17096A627), also permits the licensee to extend the IST 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 IST 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 IST 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 1 O 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 1 O CFR 50.55a(f)(5)(ii) address what to do if a revised IST 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 IST 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 IST 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, 1 O 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 5.5.6 from the TSs and renumbers the subsequent TS programs. The licensee proposes to delete the content of TS 5.5.6, but retains the TS number, and adds the word "DELETED." The licensee did not propose to renumber the subsequent TS programs. 2. The CNP TSs include plant-specific surveillances that are not contained in the STS, and/or have numbering that does not match the STS. The licensee stated that these differences are administrative and do not affect the applicability of TSTF-545 to the CNP TSs. 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 Michigan State official was notified of the proposed issuance of the amendments on April 6, 2017. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or change the surveillance requirements. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration and there has been no public comment on such finding as published in the Federal Register on September 27, 2016 (81 FR 66307). Accordingly, the amendments meet 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 amendments.
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 amendments will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributors: S. Basturescu, NRR A. Dietrich, NRR Date of issuance: May 24, 2017
ML 17103A106 OFFICE DORL/LPL3/PM DORL/LPL3/LA DE/EPNB/BC NAME JRankin SRohrer DAiiey DATE 4/25/17 4/20/17 4/13/17 OFFICE OGC-NLO DORL/LPL3/BC DORL/LPL3/PM NAME STurk DWrona JRankin
- DATE 5/3/17 5/24/17 5/24/17