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| {{#Wiki_filter:November 4, 2021 Mr. David P. Rhoades Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555 | | {{#Wiki_filter:}} |
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| ==SUBJECT:==
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| JAMES A. FITZPATRICK NUCLEAR POWER PLANT ISSUANCE OF AMENDMENT NO. 344 RE: ADOPTION OF TSTF-545 (EPID L-2020-LLA-0268)
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| ==Dear Mr. Rhoades:==
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| The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 344 to Renewed Facility Operating License No. DPR-59 for the James A.
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| FitzPatrick Nuclear Power Plant. The amendment consists of changes to the technical specifications (TS) in response to your application dated December 11, 2020, as supplemented by {{letter dated|date=May 17, 2021|text=letter dated May 17, 2021}}.
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| The amendment revises the TS to adopt Technical Specifications Task Force (TSTF) Standard Technical Specifications Change Traveler TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing, dated October 21, 2015. Specifically, the amendment removes TS 5.5.7, Inservice Testing Program, and adds a new defined term, INSERVICE TESTING PROGRAM, to TS 1.1, Definitions, and makes corresponding edits throughout the TS.
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| A copy of our related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
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| Sincerely,
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| /RA/
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| Justin C. Poole, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-333
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| ==Enclosures:==
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| : 1. Amendment No. 344 to DPR-59
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| : 2. Safety Evaluation cc: Listserv
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| EXELON FITZPATRICK, LLC AND EXELON GENERATION COMPANY, LLC DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 344 Renewed Facility Operating License No. DPR-59
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| : 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
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| A. The application for amendment by Exelon FitzPatrick, LLC and Exelon Generation Company, LLC (collectively, the licensees) dated December 11, 2020, as supplemented on May 17, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.
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| Enclosure 1
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| : 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-59 is hereby amended to read as follows:
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| (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 344, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
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| : 3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days.
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| FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by James G. James G. Danna Date: 2021.11.04 Danna 12:06:59 -04'00' James G. Danna, Chief Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
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| ==Attachment:==
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| Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: November 4, 2021
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| ATTACHMENT TO LICENSE AMENDMENT NO. 344 JAMES A. FITZPATRICK NUCLEAR POWER PLANT RENEWED FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Replace the following page of the License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
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| Remove Page Insert Page Page 3 Page 3 Replace the following pages of the 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.
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| Remove Pages Insert Pages 1.1-3 1.1-3 3.1.7-3 3.1.7-3 3.1.8-2 3.1.8-2 3.4.3-2 3.4.3-2 3.5.1-5 3.5.1-5 3.5.1-7 3.5.1-7 3.6.1.3-8 3.6.1.3-8 3.6.1.3-9 3.6.1.3-9 3.6.1.6-2 3.6.1.6-2 3.6.1.7-2 3.6.1.7-2 3.6.1.9-2 3.6.1.9-2 3.6.2.3-2 3.6.2.3-2 5.5-6 5.5-6 5.5-7 5.5-7 5.5-8 5.5-8 5.5-9 5.5-9 5.5-10 5.5-10 5.5-11 5.5-11 5.5-12 5.5-12 5.5-13 5.5-13 5.5-14 5.5-14 5.5-15 5.5-15
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| (4) Exelon Generation Company pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, at any time, any byproduct, source, and special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus, components or tools.
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| (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.
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| 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:
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| 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:
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| (1) Maximum Power Level Exelon Generation Company is authorized to operate the facility at steady state reactor core power levels not in excess of 2536 megawatts (thermal).
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| (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 344, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
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| (3) Fire Protection Exelon Generation Company shall implement and maintain in effect all provisions of the approved fire protections program as described in the Final Safety Analysis Report for the facility and as approved in the SER dated November 20, 1972; the SER Supplement No. 1 dated February 1, 1973; the SER Supplement No. 2 dated October 4, 1974; the SER dated August 1, 1979; the SER Supplement dated October 3, 1980; the SER Supplement dated February 13, 1981; the NRC Letter dated February 24, 1981; Technical Specification Amendments 34 (dated January 31, 1978), 80 (dated May 22, 1984), 134 (dated July 19, 1989),
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| 135 (dated September 5, 1989), 142 (dated October 23, 1989), 164 (dated August 10, 1990), 176 (dated January 16, 1992), 177 (dated February 10, 1992), 186 (dated February 19, 1993), 190 (dated June 29, 1993), 191 (dated July 7, 1993), 206 (dated February 28, 1994), and 214 (dated June 27, 1994); and NRC Exemptions and associated safety evaluations dated April 26, 1983, July 1, 1983, January 11, 1985, Amendment 344 Renewed License No. DPR-59
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| Definitions 1.1 1.1 Definitions (continued)
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| INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee program PROGRAM that fulfills the requirements of 10 CFR 50.55a(f).
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| ISOLATION The ISOLATION INSTRUMENTATION RESPONSE TIME shall INSTRUMENTATION be that time interval from when the monitored parameter RESPONSE TIME exceeds its isolation initiation setpoint at the channel sensor until the isolation valve receives the isolation signal (e.g., de-energization of the main steam isolation valve solenoids). 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.
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| LEAKAGE LEAKAGE shall be:
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| : a. Identified LEAKAGE
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| : 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
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| : 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:
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| : b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;
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| : c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE;
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| : d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.
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| (continued)
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| JAFNPP 1.1-3 Amendment 344
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| SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)
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| SURVEILLANCE FREQUENCY SR 3.1.7.5 Verify the concentration of sodium pentaborate in In accordance solution is within the limits of Figure 3.1. 7-1. with the Surveillance Frequency Control Program AND Once within 24 hours after water or sodium pentaborate is added to solution AND Once within 24 hours after solution temperature is restored within the limits of Figure 3.1.7-2 SR 3.1.7.6 Verify each SLC subsystem manual valve in the In accordance flow path that is not locked, sealed, or otherwise with the secured in position is in the correct position, or Surveillance can be aligned to the correct position. Frequency Control Program SR 3.1.7.7 Verify each pump develops a flow rate 50 gpm at In accordance a discharge pressure 1275 psig. with the INSERVICE TESTING PROGRAM SR 3.1.2.8 Verify flow through one SLC subsystem from pump In accordance into reactor pressure vessel. with the Surveillance Frequency Control Program (continued)
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| JAFNPP 3.1.7-3 Amendment 344
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| SDV Vent and Drain Valves 3.1.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 - - - - - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - - - - -
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| Not required to be met on vent and drain valves closed during performance of SR 3.1.8.2.
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| Verify each SDV vent and drain valve is open. In accordance with the Surveillance Frequency Control Program SR 3.1.8.2 Cycle each SDV vent and drain valve to the fully In accordance closed and fully open position. with the INSERVICE TESTING PROGRAM SR 3.1.8.3 Verify each SDV vent and drain valve: In accordance with the
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| : a. Closes in 30 seconds after receipt of an Surveillance actual or simulated scram signal; and Frequency Control Program
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| : b. Opens when the actual or simulated scram signal is reset.
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| JAFNPP 3.1.8-2 Amendment 344
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| S/RVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoint of the In accordance required S/RVs is 1145 +/- 34.3 psig. Following with the testing, lift settings shall be within +/- 1%. INSERVICE TESTING PROGRAM SR 3.4.3.2 Verify each required S/RV is capable of being In accordance opened. with the INSERVICE TESTING PROGRAM JAFNPP 3.4.3-2 Amendment 344
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| ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
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| SURVEILLANCE FREQUENCY SR 3.5.1.6 - - - - - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - - - - -
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| Not required to be performed if performed within the previous 31 days.
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| Verify each recirculation pump discharge valve Once each cycles through one complete cycle of full travel or startup prior to is de-energized in the closed position. exceeding 25%
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| RTP 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 pressure INSERVICE above primary containment pressure. TESTING PROGRAM SYSTEM HEAD CORRESPONDING TO A REACTOR PRESSURE ABOVE NO. PRIMARY OF CONTAINMENT SYSTEM FLOW RATE PUMPS PRESSURE OF Core Spray 4265 gpm 1 113 psi LPCI 7700 gpm 1 20 psi SR 3.5.1.8 - - - - - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - - - - -
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| Not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test.
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| Verify, with reactor pressure 1040 psig and In accordance 970 psig, the HPCI pump can develop a flow rate with the 3400 gpm against a system head corresponding INSERVICE to reactor pressure. TESTING PROGRAM (continued)
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| JAFNPP 3.5.1-5 Amendment 344
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| ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
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| SURVEILLANCE FREQUENCY SR 3.5.1.13 Verify each required ADS valve is capable of being In accordance opened. with the INSERVICE TESTING PROGRAM JAFNPP 3.5.1-7 Amendment 344
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| PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
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| SURVEILLANCE FREQUENCY SR 3.6.1.3.3 - - - - - - - - - - - - - - - - - - - - - - - NOTES- - - - - - - - - - - - - - - - - - - - -
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| : 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
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| : 2. Not required to be met for PCIVs that are open under administrative controls.
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| Verify each primary containment manual isolation Prior to entering valve and blind flange that is located inside MODE 2 or 3 from primary containment and not locked, sealed or MODE 4 if otherwise secured and is required to be closed primary during accident conditions is closed. containment was de-inerted while in MODE 4, if not performed within the previous 92 days SR 3.6.1.3.4 Verify continuity of the traversing incore probe In accordance (TIP) shear isolation valve explosive charge. with the Surveillance Frequency Control Program SR 3.6.1.3.5 Verify the isolation time of each power operated, In accordance automatic PCIV, except for MSIVs, is within limits. with the INSERVICE TESTING PROGRAM*
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| SR 3.6.1.3.6 Verify the isolation time of each MSIV is 3 In accordance seconds and 5 seconds. with the INSERVICE TESTING PROGRAM (continued)
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| * This Surveillance for ST-24J is not required to be performed until following the return of the A RHR pump to OPERABLE. This past due Surveillance will be completed as stated in Section 3.5 of letter JAFP-21-0053, dated June 14, 2021.
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| JAFNPP 3.6.1.3-8 Amendment 344
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| PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
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| SURVEILLANCE FREQUENCY SR 3.6.1.3.7 Verify each automatic PCIV actuates to the In accordance isolation position on an actual or simulated with the isolation signal. Surveillance Frequency Control Program SR 3.6.1.3.8 Verify each reactor instrumentation line EFCV In accordance actuates to the isolation position on a simulated with the instrument line break. INSERVICE TESTING PROGRAM SR 3.6.1.3.9 Remove and test the explosive squib from each In accordance shear isolation valve of the TIP System. with the Surveillance Frequency Control Program SR 3.6.1.3.10 Verify combined main steam line leakage rate is In accordance 200 scfh, and 100 scfh for any one steam line, with the Primary when tested at 25 psig. Containment Leakage Rate Testing Program SR 3.6.1.3.11 Verify the leakage rate of each air operated In accordance testable check valve associated with the LPCI and with the Primary CS Systems vessel injection penetrations is within Containment limits. Leakage Rate Testing Program JAFNPP 3.6.1.3-9 Amendment 344
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| Reactor Building-to-Suppression Chamber Vacuum Breakers 3.6.1.6 ACTIONS (continued)
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| CONDITION REQUIRED ACTION COMPLETION TIME D Two lines with one or D.1 Restore all vacuum 1 hour more reactor building- breakers in one line to-suppression to OPERABLE status.
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| chamber vacuum breakers inoperable for opening.
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| E. Required Action and E.1 Be in MODE 3. 12 hours Associated Completion Time not AND met.
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| E.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.6.1 - - - - - - - - - - - - - - - - - - - - - - - NOTES- - - - - - - - - - - - - - - - - - - - -
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| : 1. Not required to be met for vacuum breakers that are open during Surveillances.
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| : 2. Not required to be met for vacuum breakers open when performing their intended function.
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| Verify each vacuum breaker is closed. In accordance with the Surveillance Frequency Control Program SR 3.6.1.6.2 Perform a functional test of each vacuum breaker. In accordance with the INSERVICE TESTING PROGRAM (continued)
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| JAFNPP 3.6.1.6-2 Amendment 344
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| Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.7.1 - - - - - - - - - - - - - - - - - - - - - - - NOTES- - - - - - - - - - - - - - - - - - - - -
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| : 1. Not required to be met for vacuum breakers that are open during Surveillances.
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| : 2. Not required to be met for vacuum breakers open when performing their intended function.
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| Verify each vacuum breaker is closed. In accordance with the Surveillance Frequency Control Program SR 3.6.1.7.2 Perform a functional test of each vacuum breaker. In accordance with the INSERVICE TESTING PROGRAM SR 3.6.1.7.3 Verify the opening setpoint of each vacuum In accordance breaker is 0.5 psid. with the Surveillance Frequency Control Program JAFNPP 3.6.1.7-2 Amendment 344
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| RHR Containment Spray System 3.6.1.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.9.1 Verify each RHR containment spray subsystem In accordance manual, power operated, and automatic valve in with the the flow path that is not locked, sealed, or Surveillance otherwise secured in position, is in the correct Frequency Control position or can be aligned to the correct position. Program SR 3.6.1.9.2 Verify each required RHR pump develops a flow In accordance rate of 7750 gpm through the associated heat with the exchanger while operating in the suppression pool INSERVICE cooling mode. TESTING PROGRAM SR 3.6.1.9.3 Verify each spray nozzle is unobstructed. In accordance with the Surveillance Frequency Control Program JAFNPP 3.6.1.9-2 Amendment 344
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| RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling In accordance subsystem manual, power operated, and with the automatic valve in the flow path that is not Surveillance locked, sealed, or otherwise secured in position, Frequency Control is in the correct position or can be aligned to the Program correct position.
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| SR 3.6.2.3.2 Verify each required RHR pump develops a flow In accordance rate of 7700 gpm through the associated heat with the exchanger while operating in the suppression pool INSERVICE cooling mode. TESTING PROGRAM JAFNPP 3.6.2.3-2 Amendment 344
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| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.6 Primary Containment Leakage Rate Testing Program (continued)
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| : d. The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.
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| : e. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.
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| 5.5.7 Not Used 5.5.8 Ventilation Filter Testing Program (VFTP)
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| This program implements the following required testing of Engineered Safeguards filter ventilation systems.
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| The tests described in Specifications 5.5.8.a and 5.5.8.b shall be performed:
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| Once per 24 months; After each complete or partial replacement of the HEPA filter train or charcoal adsorber filter or after removal of a charcoal sample; After any structural maintenance on the HEPA filter or charcoal adsorber housing that could affect the filter system efficiency; and Following painting. fire, or chemical release that could adversely affect the ability of the filter system to perform the intended function in any ventilation zone communicating with the system.
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| The tests described in Specification 5.5.8.c shall be performed:
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| Once per 24 months; After 720 hours of system operation; After any structural maintenance on the charcoal adsorber housing that could affect the filter system efficiency; and Following painting, fire. or chemical release that could adversely affect the ability of the charcoal filter system to perform the intended function in any ventilation zone communicating with the system.
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| The tests described in Specifications 5.5.8.d and 5.5.8.e shall be performed once per 24 months.
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| (continued)
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| JAFNPP 5.5-6 Amendment 344
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| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Ventilation Filter Testing Program (VFTP) (continued)
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| The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test Frequencies.
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| : a. Demonstrate for each of the Engineered Safeguards systems that an inplace test of the HEPA filters shows a penetration and system bypass less than the value specified below when tested in accordance with Sections C.5.a and C.5.c of Regulatory Guide 1.52, Revision 2, and ASME N510-1980 at the system flowrate specified below.
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| Engineered Safeguards Penetration and Flowrate (scfm)
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| Ventilation System System Bypass Standby Gas 1.5% 5,400 to 6,600 Treatment System Control Room 1.5% 900 to 1,100 Emergency Ventilation Air Supply System
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| : b. Demonstrate for each of the Engineered Safeguards systems that an inplace test of the charcoal adsorber shows a penetration and system bypass less than the value specified below when tested in accordance with Sections C.5.a and C.5.d of Regulatory Guide 1.52, Revision 2, and ASME N510-1980 at the system flowrate specified below.
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| Engineered Safeguards Penetration and Flowrate (scfm)
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| Ventilation System System Bypass Standby Gas 1.0% 5,400 to 6,600 Treatment System Control Room 0.5% 900 to 1,100 Emergency Ventilation Air Supply System (continued)
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| JAFNPP 5.5-7 Amendment 344
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| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Ventilation Filter Testing Program (VFTP) (continued)
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| : c. Demonstrate for each of the Engineered Safeguards systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Section C.6.b of Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in series, in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and the relative humidity specified below.
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| Engineered Safeguards Penetration RH Ventilation System Standby Gas 1.5% 70%
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| Treatment System Control Room 1.5% 95%
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| Emergency Ventilation Air Supply System
| |
| : d. Demonstrate for each of the Engineered Safeguards systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified below:
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| Engineered Safeguards Delta P Flowrate (scfm)
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| Ventilation System (inches wg)
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| Standby Gas 5.7 5,400 to 6,600 Treatment System Control Room 5.8 900 to 1,100 Emergency Ventilation Air Supply System
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| : e. Demonstrate that the heaters for the Standby Gas Treatment System dissipate > 29 kW when tested in accordance with ASME N510-1975.
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| (continued)
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| JAFNPP 5.5-8 Amendment 344
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| Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Main Condenser Offgas Treatment System and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.
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| The program shall include:
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| : a. The limits for concentrations of hydrogen and oxygen in the Main Condenser Offgas Treatment System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and
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| : b. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste System is less than the amount that would result in a concentration that is 10 times the concentration values in Appendix B. Table 2, Column 2, to 10 CFR 20.1001*20.2402 (excluding tritium and dissolved or entrained noble gases) at the nearest potable water supply and the nearest surface water supply beyond the site boundary, in the event of an uncontrolled release of the tanks' contents.
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| The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program Surveillance Frequencies.
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| (continued)
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| JAFNPP 5.5-9 Amendment 344
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| Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.10 Diesel Fuel Oil Testing Program This program implements required testing of both new fuel oil and stored fuel oil. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:
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| : a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:
| |
| : 1. An API gravity or an absolute specific gravity within limits,
| |
| : 2. A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and
| |
| : 3. A clear and bright appearance with proper color or water and sediment within limits;
| |
| : b. Within 31 days following addition of the new fuel oil to storage tanks verify that the properties of the new fuel oil, other than those addressed in Specification 5.5.10.a above, are within limits for ASTM 2D fuel oil; and
| |
| : c. Total particulate concentration of the fuel oil is 10 mg/I when tested every 31 days in accordance with the applicable ASTM Standard, except that the specified filters may be replaced with filters up to 3.0 microns.
| |
| The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test Frequencies.
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| (continued)
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| JAFNPP 5.5-10 Amendment 344
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.11 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.
| |
| : a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
| |
| : b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
| |
| : 1. A change in the TS incorporated in the license; or
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| : 2. A change to the UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
| |
| : c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.
| |
| : d. Proposed changes that meet the criteria of Specification 5.5.11.b above shall be reviewed and approved by the NRC prior to implementation.
| |
| Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
| |
| (continued)
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| JAFNPP 5.5-11 Amendment 344
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.12 Safety Function Determination Program (SFDP)
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| This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6.
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| : a. The SFDP shall contain the following:
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| : 1. Provisions for cross division checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;
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| : 2. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists:
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| : 3. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and
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| : 4. Other appropriate limitations and remedial or compensatory actions.
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| : b. A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power or no concurrent loss of emergency diesel generator subsystems, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:
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| : 1. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or
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| : 2. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or
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| : 3. A required system redundant to support system(s) for the supported systems (1) and (2) above is also inoperable.
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| The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.
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| (continued)
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| JAFNPP 5.5-12 Amendment 344
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| Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.13 Configuration Risk Management Program (CRMP)
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| The CRMP provides a proceduralized risk-informed assessment to manage the risk associated with equipment inoperability. The program applies to technical specification structures. systems, or components for which a risk-informed allowed outage time has been granted. The program is to include the following:
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| : a. Provisions for the control and implementation of a Level 1 at-power internal events PRA-informed methodology. The assessment is to be capable of evaluating the applicable plant configuration.
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| : b. Provisions for performing an assessment prior to entering the plant configuration described by the Limiting Condition for Operation (LCO)
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| Condition(s) for preplanned activities.
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| : c. Provisions for performing an assessment after entering the plant configuration described by the LCO Condition(s) for unplanned entry into the LCO Condition(s).
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| : d. Provisions for assessing the need for additional actions after the discovery of additional equipment-out-of-service conditions while in the plant configuration described by the LCO Condition(s).
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| : e. Provisions for considering other applicable risk-significant contributors such as Level 2 issues and external events, qualitatively or quantitatively.
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| (continued)
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| JAFNPP 5.5-13 Amendment 344
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| Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.14 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation Air Supply (CREVAS) System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident. The program shall include the following elements:
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| : a. The definition of the CRE and the CRE boundary.
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| : b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.
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| : c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.
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| : d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the isolate mode of operation by one subsystem of the CREVAS System, operating at the flow rate required by the VFTP, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.
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| : e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.
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| : f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.
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| (continued)
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| JAFNPP 5.5-14 Amendment 344
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| Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.15 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
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| : a. The Surveillance Frequency Control Program shall contain a list of Frequencies of the Surveillance Requirements for which the Frequency is controlled by the program.
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| : b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies, Revision 1.
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| : c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.
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| JAFNPP 5.5-15 Amendment 344
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| SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 344 EXELON FITZPATRICK, LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT TO RENEWED FACILITY OPERATING LICENSE NO. DPR-59
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| ==1.0 INTRODUCTION==
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| By application dated December 11, 2020, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20346A025), as supplemented by {{letter dated|date=May 17, 2021|text=letter dated May 17, 2021}} (ADAMS Accession No. ML21137A005), Exelon Generation Company, LLC (Exelon, the licensee) requested changes to the technical specifications (TS) for James A.
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| FitzPatrick Nuclear Power Plant (FitzPatrick). Specifically, the licensee requested changes to the TSs consistent with Technical Specifications Task Force (TSTF) Standard Technical Specifications (STSs) Change Traveler TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing, dated October 21, 2015 (ADAMS Accession No. ML15294A555).
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| The licensees proposed changes would delete FitzPatrick TS 5.5.7 Inservice Testing Program, and add a new defined term, INSERVICE TESTING PROGRAM, to TS Section 1.1, Definitions. In addition, all existing references to the Inservice Testing Program in the FitzPatrick SRs are replaced with INSERVICE TESTING PROGRAM so that the SRs refer to the new definition in lieu of the deleted program.
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| The supplemental {{letter dated|date=May 17, 2021|text=letter dated May 17, 2021}}, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC, the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register (FR) on February 23, 2021 (86 FR 11013).
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| Enclosure 2
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| ==2.0 REGULATORY EVALUATION==
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| 2.1 Description of Inservice Testing Requirements and TSTF-545 An inservice test is a test to assess the operational readiness of a structure, system, or component after first electrical generation by nuclear heat. The American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) provides requirements for inservice testing of certain components in light-water nuclear power plants. The ASME OM Code identifies the components subject to the testing (i.e., pumps, valves, pressure relief devices, and dynamic restraints), responsibilities, methods, intervals, parameters to be measured and evaluated, criteria for evaluating results, corrective actions, personnel qualification, and recordkeeping. Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(f)(4), Inservice testing standards requirement for operating plants, requires, in part, that throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves that are within the scope of the ASME OM Code must meet the inservice test requirements (except design and access provisions) set forth in the ASME OM Code and addenda that become effective subsequent to editions and addenda specified in 10 CFR 50.55a(f)(2) and (3) and that are incorporated by reference in 10 CFR 50.55a(a)(1)(iv),
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| to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulation at 10 CFR 50.55a(g)(4), Inservice inspection standards requirement for operating plants, includes inservice examination and testing requirements for dynamic restraints (snubbers) depending on the applicability of the ASME OM Code or the ASME Boiler and Pressure Vessel Code (BPV Code). The facilitys TSs also prescribe inservice testing requirements and frequencies for ASME Code Class 1, 2, and 3 components.
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| The regulation in 10 CFR 50.55a(f)(5)(ii), IST [Inservice testing] program update: Conflicting IST Code requirements with technical specifications, 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 Inservice Testing Program from the Administrative Controls section of the TSs. The TSs contain surveillances that require testing or test intervals in accordance with the Inservice Testing Program. The elimination of the Inservice 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, Inservice Testing Program, with the defined term, as denoted by capitalized letters, throughout the TSs.
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| The NRC approved TSTF-545, Revision 3, by {{letter dated|date=December 11, 2015|text=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).
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| 2.2 Proposed Technical Specifications Changes The licensee requested to delete TS 5.5.7, from the Administrative Controls section of TSs and replace it with the words Not Used. TS 5.5.7 currently reads:
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| | |
| This program provides controls for inservice testing of certain ASME Code Class 1, 2, and 3 pumps and valves. The program shall include the following:
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| : a. Testing frequencies specified ASME OM Code and applicable Addenda are as follows:
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| ASME OM Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice activities testing activities Quarterly or every 3 months At least once per 92 days Biennially or every 2 years At least once per 731 days
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| : b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;
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| : c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and
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| : d. Nothing in the ASME Boiler and Pressure Vessel code shall be construed to supersede the requirements of any TS.
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| SR 3.0.2 allows an extension of inservice testing intervals by up to 25 percent. If it is discovered that a surveillance associated with an inservice testing activity was not performed within the required interval, SR 3.0.3 allows the licensee to delay declaring the associated limiting condition for operation not met in order to perform the missed surveillance. The licensee did not request changes to SR 3.0.2 or SR 3.0.3.
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| 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 proposed that all existing occurrences of Inservice 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.
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| 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:
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| 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) surveillance requirements; (4) design features; (5) administrative controls;
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| | |
| (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.
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| The NRC staffs 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]
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| Edition, Chapter 16, Section 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 licensees proposed amendment is based on TSTF-545, Revision 3, which is an NRC-approved change to the improved STSs. The NRC staffs 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 NRC-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.
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| Inservice 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:
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| Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements for preservice and inservice testing (referred to in this paragraph (f) collectively as inservice testing) of the ASME BPV Code and ASME OM Code as specified in this paragraph (f). 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)]. . . .
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| 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 NRCs rules.
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| The applicable NRC regulations are found in 10 CFR 50.55(a)(f)(5)(ii) which addresses a plants IST program.
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| NUREG-1482, Revision 3, Guidelines for Inservice Testing at Nuclear Power Plants: Inservice Testing of Pumps and Valves and Inservice Examination and Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants, July 2020 (ADAMS Accession No. ML20202A473) provides guidance for the inservice testing of pumps, valves and dynamic restraints.
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| | |
| NUREG-0800, Section 3.9.6, Functional Design, Qualification, and Inservice Testing Programs for Pumps, Valves, and Dynamic Restraints, Revision 4, March 2017 (ADAMS Accession No. ML16134A116), provides guidance and acceptance criteria for the NRC staff review of the inservice testing program for pumps, valves and dynamic restraints.
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| ==3.0 TECHNICAL EVALUATION==
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| | |
| The NRC staff evaluated the licensees 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)
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| (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 NRC staff considered those regulatory requirements that are automatically conditions of the license through 10 CFR 50.54.
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| 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.
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| 3.1 Deletion of the Inservice Testing Program from the TSs TS 5.5.7 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).
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| 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.
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| Consideration of TS 5.5.7.a The ASME OM Code requires testing to normally be performed within certain time periods.
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| TS 5.5.7.a sets inservice testing frequencies (see also FitzPatrick Amendment No. 296, ADAMS Accession No. ML100890188) consistent with inservice test frequencies specified in the ASME OM Code and applicable addenda. As described above, requiring the licensee to have an inservice testing program in TSs is duplicative of the license condition in 10 CFR 50.54.
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| Therefore, the NRC staff determined that the specifying inservice testing frequencies in TSs is not necessary to assure operation of the facility in a safe manner.
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| Consideration of TS 5.5.7.b TS 5.5.7.b allows the licensee to extend, by up to 25 percent, the interval between inservice testing activities, as required by TS 5.5.7.a. The NRC regulations in 10 CFR 50.55a incorporate by reference certain codes, standards, and NRC regulatory guides (RGs), including RG 1.192, Revision 3, Operation and Maintenance Code Case Acceptability, ASME OM Code, in
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| 10 CFR 50.55a(b)(6). RG 1.192 lists ASME OM Code Cases that the NRC has accepted with or without conditions, including ASME OM Code Case OMN-20, Inservice Test Frequency.
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| ASME Code Case OMN-20 permits specific grace periods for inservice testing intervals required in the ASME OM Code where the provisions of Code Case OMN-20 are applied. Similar to TS 5.5.7.b, for inservice testing periods up to and including two years, the ASME Code Case OMN-20 permits the licensee to extend the inservice testing intervals specified in the ASME OM Code by up to 25 percent.
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| The NRC staff determined that the TS 5.5.7.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 5.5.7.b is acceptable. The deletion of TS 5.5.7.b does not impact the licensees ability to extend inservice testing intervals using Code Case OMN-20, in accordance with 10 CFR 50.55a.
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| Consideration of TS 5.5.7.c TS 5.5.7.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.
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| Deletion of TS 5.5.7.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 5.5.7.c is acceptable.
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| Consideration of TS 5.5.7.d TS 5.5.7.d states that nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS. However, the NRC 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; the regulations 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.
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| Conclusion Regarding Deletion of TS 5.5.7 The NRC staff determined that the requirements currently in TS 5.5.7 are not necessary to assure operation of the facility in a safe manner. Based on this evaluation, the NRC staff concludes that deletion of TS 5.5.7 from the licensees TSs is acceptable, because TS 5.5.7 is not required by 10 CFR 50.36(c)(5).
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| 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,
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| 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).
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| The licensee proposed that all existing references to the Inservice 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 Inservice 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. Based on its review, the NRC 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 NRC staff also determined that 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.
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| 3.3 Variations from TSTF-545 In its application, the licensee identified the following variations from TSTF-545, Revision 3:
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| TSTF-545, Revision 3, completely deletes TS 5.5.7 from the TSs and renumbers the subsequent TS programs. The licensee proposed to delete TS 5.5.7 from the Administrative Controls section of TSs, keep the number and add the word Not Used.
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| The licensee did not propose to renumber the subsequent TS programs.
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| Some of the SRs modified in TSTF-545, Revision 3, are not applicable to the licensee.
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| Some of the SRs modified in TSTF-545, Revision 3, are applicable to the licensee however, the licensees uses a different SR number.
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| Some licensee plant-specific SRs (not identified in TSTF-545, Revision 3) have an SR frequency as In accordance with the Inservice Testing Program. These SRs will be changed to In accordance with the INSERVICE TESTING PROGRAM.
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| The NRC staff finds that the proposed variations are editorial in nature and the licensees proposed TS changes remain consistent with the intent of TSTF-545, Revision 3. Therefore, the staff finds that the licensees proposed TS changes are acceptable.
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| ==4.0 STATE CONSULTATION==
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| In accordance with the Commissions regulations, the New York State official was notified of the proposed issuance of the amendment on August 10, 2021. The State official had no comments.
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| ==5.0 ENVIRONMENTAL CONSIDERATION==
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| | |
| The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public
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| | |
| comment on such finding (February 23, 2021; 86 FR 11013). 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.
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| ==6.0 CONCLUSION==
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| 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.
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| Principal Contributor(s): C. Ashley Date of Issuance: November 4, 2021
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| ML21181A057 *via memorandum OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DSS/STSB/BC*
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| NAME JPoole KZeleznock VCusumano DATE 08/11/2021 08/10/2021 05/24/2021 OFFICE NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME JDanna JPoole DATE 11/04/2021 11/04/2021}}
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