ML20024C661

Issuance of Amendment No. 332 Adopt TSTF-568, Revision 2, Revise Applicability of BWR/4 TS 3.6.2.5 and TS 3.6.3.2, Using the Consolidated Line Item Improvement Process

ML20024C661
Person / Time
Site:	FitzPatrick
Issue date:	03/02/2020
From:	Samson Lee Plant Licensing Branch 1
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
Lee S, NRR/DORL/LPL1, 415-3168
References
EPID L-2020-LLA-0008
Download: ML20024C661 (22)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 March 2, 2020 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT- ISSUANCE OF AMENDMENT NO. 332 RE: ADOPT TSTF-568, REVISION 2, "REVISE APPLICABILITY OF BWR/4 TS 3.6.2.5 AND TS 3.6.3.2," USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS (EPID L-2020-LLA-0008)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 332 to Renewed Facility Operating License No. DPR-59 for the James A.

FitzPatrick Nuclear Power Plant (FitzPatrick) in response to your application dated January 23, 2020 (Agencywide Documents Access and Management System Accession No. ML20023A362). Your application requested the adoption of Technical Specifications Task Force (TSTF) Traveler TSTF-568, Revision 2, "Revise Applicability of BWR [Boiling Water Reactor]/4 TS [Technical Specification] 3.6.2.5 and TS 3.6.3.2," using the Consolidated Line Item Improvement Process.

The amendment revises FitzPatrick TS 3.6.2.4, "Drywell-to-Suppression Chamber Differential Pressure," and TS 3.6.3.1, "Primary Containment Oxygen Concentration," and presents the requirements in a manner more consistent with the Standard Technical Specifications format and content.

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

Sincerely, Samson S. Lee, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-333

Enclosures:

1. Amendment No. 332 to DPR-59

2. Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 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. 332 Renewed Facility Operating License No. DPR-59

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

A. The application for amendment by Exelon FitzPatrick, LLC and Exelon Generation Company, LLC (collectively, the licensees) dated January 23, 2020, 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.

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:

Enclosure 1

2. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 332, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION

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James G. Danna, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: March 2, 2020

ATTACHMENT TO LICENSE AMENDMENT NO. 332 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 a marginal line indicating the area of change.

Remove Page Insert Page 3 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.

Remove Pages Insert Pages 3.6.2.4-1 3.6.2.4-1 3.6.3.1-1 3.6.3.1-1

(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.

(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 Exelon Generation Company is authorized to operate the facility at steady state reactor core power levels not in excess of 2536 megawatts

{thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 332, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

(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),

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 332 Renewed License No. DPR-59

Drywell-to-Suppression Chamber Differential Pressure 3.6.2.4 3.6 CONTAINMENT SYSTEMS 3.6.2.4 Drywell-to-Suppression Chamber Differential Pressure LCO 3.6.2.4 The drywell pressure shall be maintained ~ 1. 7 psi above the pressure of the suppression chamber.

APPLICABILITY: MODE 1 with THERMAL POWER > 15% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell-to-suppression A.1 ---------NOTE---------- 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> chamber differential LCO 3.0.4.c is Applicable.

pressure not within -------------------------

limit.

Restore differential pressure to within limit.

B. Required Action and B.1 Reduce THERMAL 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion POWER to~ 15% RTP.

Time not met.

JAFNPP 3.6.2.4-1 Amendment 332

Primary Containment Oxygen Concentration 3.6.3.1 3.6 CONTAINMENT SYSTEMS 3.6.3.1 Primary Containment Oxygen Concentration LOO 3.6.3.1 The primary. containment oxygen concentration shall be < 4.0 volume percent APPLICABILITY:. MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary containment A.1 ---------NOTE---------- 72 hou.-s oxygen concentration LCO 3.0.4.c is Applicable.

not within limit -------------------------

Restore oxygen concentration to within limit.

B. Required Action and B.1 Bein MODE3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1.1 Verify primary containment oxygen concentration In accordance with Is within limits. the Surveillance Frequency Control Program JAFNPP 3.6.3.1-1 Amendment 332

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 332 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-59 EXELON FITZPATRICK, LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT

1.0 INTRODUCTION

By application dated January 23, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20023A362), Exelon Generation Company, LLC (the licensee) submitted a license amendment request (LAR) for the James A. FitzPatrick Nuclear Power Plant (FitzPatrick).

The proposed changes would revise FitzPatrick Technical Specification (TS) 3.6.2.4, "Drywell-to-Suppression Chamber Differential Pressure," and TS 3.6.3.1, "Primary Containment Oxygen Concentration." Specifically, the proposed changes would simplify and clarify the applicability statements, which if misapplied, could conflict with the corresponding required actions. The proposed changes would also remove the undefined term "scheduled reactor shutdown" and provide adequate terminal actions.

The proposed amendment is based on Technical Specifications Task Force (TSTF) Traveler TSTF-568, Revision 2, "Revise Applicability of BWR [Boiling Water Reactor]/4 TS 3.6.2.5 and TS 3.6.3.2" (TSTF-568, Revision 2, or the traveler) (ADAMS Accession No. ML19141A122).

The U.S. Nuclear Regulatory Commission {NRC, the Commission) approved TSTF-568, Revision 2, by letter dated December 17, 2019 (ADAMS Package Accession No. ML19325C444). The NRC staff's safety evaluation (SE) of the traveler was enclosed with the NRC staff's approval letter.

The licensee is not proposing any variations from the TS changes described in TSTF-568, Revision 2, or the applicable parts of the NRC staff's SE of TSTF-568, Revision 2. The only difference between the licensee's submittal and TSTF-568, Revision 2, is the numbering of the TSs. Specifically, the FitzPatrick TS numbers are incrementally one lower than those in the traveler such that TS 3.6.2.5 in the traveler is 3.6.2.4 for FitzPatrick and TS 3.6.3.2 in the traveler is 3.6.3.1 for FitzPatrick. The TS numbers in this SE reflect the FitzPatrick TS numbers except where quoting TS numbers from the traveler.

Enclosure 2

2.0 REGULATORY EVALUATION

2.1 Description of Structures, Systems, and Components, and TS Sections 2.1.1 Current Drywell-to-Suporession Chamber Differential Pressure Control The drywell-to-suppression chamber differential pressure control is a safety-related operational feature of Mark I containment designs. The TS 3.6.2.4 requires a minimum differential pressure of 1.7 pounds per square inch differential (psid) to reduce the loss-of-coolant accident (LOCA) hydrodynamic loads during the Mark I containment load definition short- and long-term programs. 1 The LOCA pool swell loads are significantly reduced because the differential pressure control reduces the length of water leg in the downcomer. The LOCA vent clearing and pool swell due to bubble formation would occur earlier (i.e., at a lower drywell pressure resulting in lesser forces on the suppression chamber, thereby increasing the safety margin for containment integrity, containment internal structures, and pressure boundary). Decreasing the allowable suppression chamber water level has a similar effect.

It is difficult to control the differential pressure during startup and shutdown transients. This is because of the variation of the drywell heat loads from the primary and auxiliary systems and because the inerting (during startup) or the de-inerting (during shutdown) of containment.

lnerting the containment during startup involves the addition of large volumes of nitrogen.

De-inerting containment during shutdown involves the addition of large volumes of air. In order to allow operation during the time differential pressure control is difficult, the current TS 3.6.2.4 is applicable from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following startup after the reactor thermal power exceeds 15 percent reactor thermal power (RTP) to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing thermal power less than 15 percent RTP during the next scheduled reactor shutdown.

2.1.2 Current Primary Containment Oxygen Concentration Requirement The regulation at Title 10 of the Code of Federal Regulations ( 10 CFR) Section 50.44, "Combustible gas control for nuclear power reactors," states that for a plant with an inerted containment atmosphere, the oxygen concentration in the primary containment is required to be maintained below 4 percent by volume during normal plant operation. This requirement ensures that an accident that produces hydrogen does not result in a combustible mixture inside the primary containment. The current TS 3.6.3.1 requires primary containment oxygen concentration to be less than 4 percent by volume when in Mode 1 during the period from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the thermal power exceeds 15 percent RTP following startup to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing thermal power to less than 15 percent RTP during the next scheduled reactor shutdown. TSTF-568, Revision 2, stated that the 24-hour allowance above 15 percent RTP is provided in the primary containment oxygen concentration specification to delay inerting the primary containment in a plant startup and to accelerate de-inerting for a plant shutdown. This allowance is provided so that plant personnel can safely enter the primary containment without breathing apparatus to perform the needed inspections and maintenance adjustments.

The containment consists of a drywell (in the shape of an inverted light bulb), a suppression chamber (in the shape of a toroid}, and a network of vents that extends radially outward and downward from the drywell to the suppression chamber. The containment atmosphere is 1 U.S. Nuclear Regulatory Commission, NUREG-0661, "Safety Evaluation Report, Mark I Containment Long-Term Program, Resolution of Generic Technical Activity A-7," dated July 1980 (ADAMS Accession No. ML072710452).

inerted with nitrogen gas during normal operation to prevent a combustible mixture of hydrogen and oxygen from forming during accident conditions. Long-term control of post-LOCA hydrogen gas concentration is accomplished by adding additional nitrogen gas and then venting the primary containment through the standby gas treatment system.

2.1.3 Pressure Suppression Following a LOCA The drywell is immediately pressurized when a postulated line break occurs within the primary containment. As drywell pressure increases, drywell atmosphere (primarily nitrogen gas) and steam are blown down through the vents into the suppression pool via the downcomers. The steam condenses in the suppression pool, which suppresses the peak pressure in the drywell.

Non-condensable gases discharged into the suppression pool collect in the free air volume of the suppression chamber, increasing the suppression chamber pressure. As steam is condensed in the suppression pool and on the structures in the drywell, the pressure decreases until the suppression chamber pressure exceeds the drywell pressure and the suppression chamber-drywell vacuum breakers open and vent non-condensable gases back into the drywall.

2.1.4 TS 3.6.2.4, "Drywell-to-Suppression Chamber Differential Pressure" A drywell-to-suppression chamber differential pressure limit is required to ensure the containment conditions assumed in the safety analyses are met. Failure to maintain the required differential pressure could result in excessive forces on the suppression chamber due to higher water clearing loads from downcomer vents and higher-pressure buildup in the drywall during a LOCA. Drywell-to-suppression chamber differential pressure must be controlled when the primary containment is inert. The TS requires that the drywell pressure be maintained

~ 1. 7 psid above the pressure of the suppression chamber.

2.1.5 TS 3.6.3.1. "Primary Containment Oxygen Concentration" The primary containment oxygen concentration is maintained to ensure that a LOCA, a postulated event that produces hydrogen, does not result in a combustible mixture inside primary containment. The TS requires that the primary containment oxygen concentration be maintained below 4 volume percent. Below this concentration, the primary containment is inerted and no combustion can occur.

2.2 Description of Proposed TS Changes 2.2.1 Proposed Changes to TS 3.6.2.4, "Drywell-to-Suppression Chamber Differential Pressure" The Applicability of TS 3.6.2.4, "Drywall-to-Suppression Chamber Differential Pressure," would be revised as shown below.

MODE 1 during the time period:

a. From 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is > 15% RTP following startup, to

b. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown.

Required Action A.1 and its associated completion time (CT) would be revised as shown below (additions in italics and deletions in strike-through).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell-to-suppression A.1 -------------NO TE------------ 72 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> chamber differential LCO 3.0.4.c is Applicable.

pressure not within ----------------------------------

limit. Restore differential pressure to within limit.

The NRC staff understands that the overall purpose of the proposed changes is to simplify the applicability statement by adding a new note and revising the CT. This change would provide similar operational flexibility but more closely follow established TS conventions.

2.2.2 Proposed Changes to TS 3.6.3.1, "Primary Containment Oxygen Concentration" The Applicability of TS 3.6.3.1, "Primary Containment Oxygen Concentration," would be revised as shown below.

Pro osed TSA MODE 1 during the time period: MODES 1 and 2.

a. From 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is > 15% RTP following startup, to

b. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown.

Required Actions A.1 and B.1 and their associated CTs would be revised as shown below (additions in italics and deletions in strike-through).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary containment A.1 -------------NO TE------------- 72~ hours oxygen concentration LCO 3.0.4.c is Applicable.

not within limit. ----------------------------------

Restore oxygen concentration to within limit.

B. Required Action and B.1 Be in MOOE 3 Reeh:1se 12 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> associated Completion THERMAL POVVER to :S: 15%

Time riot met. R+P.

The NRC staff understands that the overall purpose of the proposed changes is to simplify the applicability statement by adding a new note and revising the CT. This change would provide operational flexibility but more closely follow established TS conventions and require that the plant be in Mode 3 if oxygen concentration cannot be restored to within limits.

2.3 Applicable Regulatory Requirements and Guidance Section 50.90 of 10 CFR, "Application for amendment of license, construction permit, or early site permit," requires that whenever a licensee desires to amend the license, application for an amendment must be filed with the Commission fully describing the changes desired, and following as far as applicable, the form prescribed for original applications.

Under 10 CFR 50.92(a), determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses or construction permits to the extent applicable and appropriate. Both the common standards for licenses and construction permits in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be "reasonable assurance" that the activities at issue will not endanger the health and safety of the public.

The regulation 10 CFR 50.36, "Technical specifications," establishes the regulatory requirements related to the content of TSs. Section 50.36(a)(1) requires an application for an operating license to include proposed TSs. A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, must also be included in the application, but shall not become part of the TSs.

The regulation 10 CFR 50.36(b) requires:

Each license authorizing operation of a ... utilization facility ... will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to [10 CFR] 50.34[, "Contents of applications; technical information."] The Commission may include such additional technical specifications as the Commission finds appropriate.

The categories of items required to be in the TSs are listed in 10 CFR 50.36(c).

In accordance with 10 CFR 50.36(c)(2), limiting conditions for operation (LCOs) are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When LCOs are not met, the licensee must shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met. In addition, 10 CFR 50.36(c)(2)(ii)(B) requires that a TS LCO of a nuclear reactor must be established for a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The regulation 10 CFR 50.44(b)(2)(i) states that "[a]II boiling water reactors with Mark I or Mark II type containments must have an inerted atmosphere." Section 50.44(a)(1) defines

"[i]nerted atmosphere" as a containment atmosphere with less than 4 percent oxygen by volume.

Chapter 6.2.1.1.C, Revision 7, "Pressure-Suppression Type BWR Containments," of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition" (SRP}, dated March 2007 (ADAMS Accession No. ML063600403} states: "The acceptability of [LOCA] pool dynamic loads for plants with Mark I containments is based on conformance with NRC acceptance criteria found in NUREG-0661."

The NRC staff's guidance for the review of TSs is in Chapter 16.0, Revision 3, "Technical Specifications," of the SRP, dated March 2010 (ADAMS Accession No. ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared Standard Technical Specifications (STSs) for each of the LWR nuclear designs. Accordingly, the NRC staff's review includes consideration of whether the proposed changes are consistent with the applicable reference STSs (i.e., the current STSs), as modified by NRG-approved travelers. The STS applicable to FitzPatrick is NUREG-1433, Revision 4.0, "Standard Technical Specifications, General Electric BWR/4 Plants," Volume 1, "Specifications," and Volume 2, "Bases," dated April 2012 (ADAMS Accession Nos. ML12104A192 and ML12104A193, respectively}.

3.0 TECHNICAL EVALUATION

The proposed changes are based on the NRG-approved TSTF-568, Revision 2. The NRC staff also considered the regulations and guidance discussed in Section 2.3 of this SE in its review.

3.1 Proposed Changes to TS 3.7.2.4 3.1.1 Proposed Changes in the Applicability The licensee proposed to delete the time periods, dependent on startup and shutdown times, from the applicability section and to replace them with a thermal power value. These time periods are "a. From 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is> 15% RTP following startup," to "b.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown." These time periods would be replaced by flexibilities and requirements in the revised completion times and the inserted note referencing LCO 3.0.4.c. This would result in requiring the drywell pressure during Mode 1 to be maintained above the specified limit whenever the thermal power is greater than 15 percent RTP. The current limitations of applicability, dependent on startup and shutdown times, were established to allow licensees operational flexibilities, such as containment entry to perform maintenance and surveillances while at power.

In TSTF-568, Revision 2, Attachment, General Electric (GE} Safety Communication (SC} 02-10, page 4, under the heading "Corrective/Preventive Actions," item 2, it is recommended that Mark I plants that use TS 3.6.2.5 should confirm that their containment is structurally designed for pool swell loads associated with a zero drywell-to-suppression chamber differential pressure.

For these plants, the Mark I containment load definition program has defined the pool swell loads associated with zero drywell-to-suppression chamber differential pressure. NUREG-0661, Appendix A, Section 2.3, states that each plant with a differential pressure control (i.e.,

TS 3.6.2.5} perform a structural assessment to demonstrate that the containment can maintain its functional capability when the differential pressure control is out-of-service (i.e., the differential pressure is zero). (Note that the TS numbers in this paragraph are not changed to the licensee's numbers since the paragraph is referring to TSTF-568. The traveler numbers are retained in this paragraph.)

FitzPatrick is applying the drywall-to-suppression chamber differential pressure control TS 3.6.2.4. The licensee's plant-specific analysis report called Plant Unique Analysis Report (PUAR) was approved by the NRC. 2 As stated in GE SC02-10, page 3, structural assessment based on zero drywall-to-suppression chamber differential pressure pool swell load definition was used to confirm the functional capability of the suppression chamber against the Service Level D limit. The GE SC02-10 also identifies the following two major conservatisms in the pool swell load definitions based on the Mark I Quarter Scale tests:

• The drywell pressurization test transient was based on the predicted drywell pressure from the NRG-approved conservative GE code M3CPT. This code predicts about 50 percent higher drywall pressurization than a realistic analysis using the GE-Hitachi code TRACG.

• The break was simulated by air to pressurize the drywell, which produces a more severe pool swell response than a realistic nitrogen/steam mixture and enhances the bubble growth.

The NRC approval of the PUAR confirmed that the licensee met the acceptance criteria specified in NUREG-0661, Appendix A, and that the NRC reviewed and approved any exceptions that the licensee may have taken from the acceptance criteria. Therefore, the NRC approval of the PUAR confirmed that with the drywall-to-suppression chamber differential pressure out-of-service, the containment is structurally designed for the pool swell loads during a large-break LOCA.

Based on the PUAR, the NRC staff finds it acceptable for the reactor to not be depressurized when the differential pressure is out-of-service at s 15 percent RTP. Further NUREG-0661, Section 3.12.7, concluded that if the differential pressure is out-of-service, the probability of occurrence of a large-break LOCA is less than 1OE-7 per reactor-year, which is sufficiently small. This minimal probability of occurrence paired with the short period during which plants are in the transition state of less than 15 percent RTP, support the adequacy of this change because the LOCA dynamic loads are not adversely affected. The NRC staff determined that the proposed deletion of the time periods is acceptable because they are now included in the note insertion (discussed in Section 3.1.2 of this SE) and the change in the CT (discussed in Section 3.1.3 of this SE). In addition, the proposed change is acceptable since it simplifies and clarifies the applicability statement and continues to provide the lowest functional capability of equipment required for safe operation of the facility as required by 10 CFR 50.36(c)(2) by protecting containment integrity.

3.1.2 Proposed Changes in Required Action A.1 In accordance with NRG-approved Traveler TSTF-568, Revision 2, the licensee proposed to add the following note to Required Action A.1: "LCO 3.0.4.c is applicable." LCO 3.0.4 states:

When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

2 Vassallo, Domenic 8., U.S. Nuclear Regulatory Commission, letter to C. A. McNeil!, Jr., Power Authority of the State of New York, "Mark I Containment Long Term Program Re: James A. Fitzpatrick Nuclear Power Plant," dated December 12, 1984 (ADAMS Accession No. ML19203A093).

a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time;

b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications; or

c. When an allowance is stated in the individual value, parameter, or other Specification.

This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

The criterion applicable to TS LCO 3.6.2.4 is LCO 3.0.4.c since this LCO establishes an individual value or parameter (i.e., drywell pressure maintained above a certain value). The new note will allow entry into the mode of applicability of TS LCO 3.6.2.4 with the drywell pressure outside of the required limit. This note allows the licensee operational flexibility as it permits entry into Mode 1 at greater than 15 percent RTP when dryweU pressure is outside of the required limit during startup configurations. The NRC staff concludes that the addition of the note is acceptable because it clarifies and simplifies the intent of the current TS LCO 3.6.2.4 applicability statement "a." of allowing startup operation with the LCO not met.

3.1.3 Proposed Changes in the CT of Condition A In accordance with NRG-approved Traveler TSTF-568, Revision 2, the licensee proposed to change the CT for Required Action A.1 from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. TSTF-568, Revision 2, stated that the proposed change will permit safe entry of personnel into the containment in Modes 1 and 2. The 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provides: 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to de-inert the containment to permit safe personnel access, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to perform the required work, and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to re-inert containment. The NRC staff finds that the extended CT incorporates the time currently allowed through the applicability statement in Section 3.1.1 of this SE. The NRC staff finds that 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is reasonable to conduct these activities based on operating experience, and the requested completion time does not present a significant change in risk given the low probability that a large line break would occur during this period. Therefore, the NRC staff finds this change acceptable.

3.1.4 Conclusion for Proposed Changes to TS 3.6.2.4 The NRC staff finds the changes proposed in TS 3.6.2.4 acceptable and that they continue to meet 10 CFR 50.36(c)(2) since the revised LCO provides the lowest functional capability of equipment required for safe operation of the facility by protecting containment integrity.

3.2 Proposed Changes to TS 3.6.3.1 3.2.1 Proposed Changes in the Applicability In accordance with NRC-approved Traveler TSTF-568, Revision 2, the licensee proposed to expand the applicability of this LCO to Modes 1 and 2 without exception. The NRC staff finds

the proposed change acceptable because it is more restrictive since an unlikely LOCA event leading to a degraded core that could produce hydrogen has the highest probability of occurrence during Modes 1 and 2 conditions.

3.2.2 Proposed Changes in Required Action A.1 In accordance with NRG-approved Traveler TSTF-568, Revision 2, the licensee proposed to add the following note to Required Action A.1: "LCO 3.0.4.c is applicable." As stated in Section 3.1.2 of this SE, TS LCO 3.0.4.c allows entering the mode of applicability of TS LCO 3.6.3.1 with the LCO not met. Therefore, the proposed change would permit entry into Modes 1 and 2 with primary containment oxygen concentration higher than the required limit.

The NRC staff concludes that the addition of the note is acceptable because it clarifies and simplifies the intent of the current TS LCO 3.6.3.1 applicability statement "a." of allowing startup operation with the LCO not met.

3.2.3 Proposed Changes in the CT of Condition A In accordance with NRG-approved Traveler TSTF-568, Revision 2, the licensee proposed changing the CT from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 72 based on the following sequence of operations: allow 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to de-inert the containment to permit safe personnel access, allow 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to perform the required maintenance or repair work, and allow 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to re-inert the containment. The NRC staff determined that the presence of a higher oxygen concentration for the 72-hour CT is appropriate, considering the low safety significance of the change for potential accidents and the additional restrictions and conservatisms in the revised applicability.

3.2.4 Proposed Changes in Required Action B.1 In accordance with NRG-approved Traveler TSTF-568, Revision 2, the licensee proposed to change the applicability statement of TS LCO 3.6.3.1 to Modes 1 and 2. If the oxygen concentration cannot be restored within the required limit and CT of Required Action A.1, the reactor should be brought to Mode 3. In this mode, the reactor would be in a hot shutdown condition (control rods fully inserted) with all reactor vessel head bolts fully tensioned.

The NRC staff recognizes that on entering Mode 3, the decay heat is rapidly decreasing. Steam is initially available for operating the reactor core isolation cooling/high pressure coolant injection steam turbine-driven pumps until the reactor pressure and thus water temperature is substantially reduced. As the decay heat continues to decrease, operators have increased time and options for achieving adequate water injection using the low-pressure emergency core cooling system to avoid core damage and associated generation of combustible gas.

Therefore, the occurrence of a LOCA leading to degraded core is highly unlikely in Mode 3.

The NRC staff finds the proposed change in Required Action 8.1 acceptable because it provides a more appropriate terminal action since it requires the plant to be placed in a mode in which the LCO does not apply and the oxygen concentration limit is no longer required. The previous terminal action allowed an indefinite period of operation at s 15 percent RTP.

Due to the low potential for hydrogen generation when the reactor is in Mode 3, inerting of containment in Mode 3 is not needed. Therefore, the NRC staff concludes that the proposed change is acceptable because it continues to protect containment integrity and meets 10 CFR 50.36(c)(2) by providing the lowest functional capability of equipment required for safe operation of the plant.

3.2.5 Proposed Changes in the CT of Condition B In accordance with NRG-approved Traveler TSTF-568, Revision 2, the licensee proposed to change the Condition B CT from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, stating that 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is a reasonable time to reduce reactor power from full power conditions to Mode 3 in an orderly manner and without challenging plant systems. The proposed change from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for bringing the reactor to a hot shutdown condition from full power is acceptable to the NRC staff because it is not a significant change and it is based on industry operating experience.

3.2.6 Conclusion for Proposed Changes to TS 3.6.3.1 The NRC staff concludes that the proposed changes in the applicability statement for TS 3.6.3.1 are acceptable since they are more restrictive as the applicability now extends to Modes 1 and 2 without exception. In addition, the occurrence of a LOCA that could lead to degraded core conditions with containment de-inerted, while in Mode 3, is unlikely. Therefore, the changes proposed in TS 3.6.3.1 are acceptable and continue to meet 10 CFR 50.36(c)(2).

3.3 Additional Changes The licensee identified differences between the TSs for FitzPatrick and NUREG-1433, upon which TSTF-568, Revision 2, is based. These differences are limited to the TS numbering as discussed in Section 1.0 of this safety evaluation. The NRC staff determined that these differences do not affect the applicability of TSTF-568, Revision 2, or the applicable parts of the NRC staff's SE of TSTF-568, Revision 2, to FitzPatrick.

4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

The NRC's regulation in 10 CFR 50.92(c) states that the NRC may make a final determination, under the procedures in 10 CFR 50.91, that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

An evaluation of the issue of no significant hazards consideration is presented below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises the Applicability and Actions of [FitzPatrick]

TS 3.6.2.4, "Drywell-to- Suppression Chamber Differential Pressure," and

[FitzPatrick] TS 3.6.3.1, "Primary Containment Oxygen Concentration," and presents the requirements in a manner more consistent with the STS format and content. Drywell-to-suppression chamber differential pressure and primary containment oxygen concentration are not initiators to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not affected by the proposed change.

Drywell-to-Suppression Chamber Differential Pressure and Primary Containment Oxygen Concentration are assumptions in the mitigation of some accidents previously evaluated. The Applicability of TS 3.6.3.1 is changed from Mode 1 when thermal power is greater than 15% to Modes 1 and 2. This expands the Applicability of the TS and will not have an effect on the consequences of an accident. The existing Applicability exceptions are removed and replaced with a longer Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The consequences of an event that could affect the drywell-to-suppression chamber differential pressure and primary containment oxygen concentration are no different during the proposed Completion Time than the consequences of the same event during the existing Completion Times. A note referencing Limiting Condition for Operation (LCO) 3.0.4.c is added to the Actions to permit entering the Applicability with the LCO not met. The note replaces the existing Applicability exceptions. This change is administrative and has no effect on the consequences of an accident.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change revises the Applicability and Actions of [FitzPatrick]

TS 3.6.2.4, "Drywell-to- Suppression Chamber Differential Pressure," and

[FitzPatrick] TS 3.6.3.1, "Primary Containment Oxygen Concentration," and presents the requirements in a manner more consistent with the STS format and content.

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). No credible new failure mechanisms, malfunctions, or accident initiators that would have been considered a design basis accident in the UFSAR [Updated Final Safety Analysis Report] are created because the Nuclear Regulatory Commission has determined that hydrogen generation is not risk significant for design basis accidents.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change revises the Applicability and Actions of [FitzPatrick]

TS 3.6.2.4, "Drywell-to- Suppression Chamber Differential Pressure," and

[FitzPatrick] TS 3.6.3.1, "Primary Containment Oxygen Concentration," and presents the requirements in a manner more consistent with the STS format and content. No safety limits are affected. No Limiting Conditions for

Operation or Surveillance limits are affected. The Drywall-to-Suppression Chamber Differential Pressure and Primary Containment Oxygen Concentration Technical Specification requirements assure sufficient safety margins are maintained, and that the design, operation, surveillance methods, and acceptance criteria specified in applicable codes and standards (or alternatives approved for use by the NRG) will continue to be met as described in the plant's licensing basis. The proposed change does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluation, the NRG staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRG staff has made a final determination that no significant hazards consideration is involved for the proposed amendment and that the amendment should be issued as allowed by the criteria contained in 10 CFR 50.91.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment on January 24, 2020. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRG 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, which was published in the Federal Register on January 29, 2020 (85 FR 5256), and there has been no public comment on such finding. 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.

7 .0 CONCLUSION The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the

amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: S. Smith Date: March 2, 2020

8. Hanson

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT NO. 332 RE: ADOPT TSTF-568, REVISION 2, "REVISE APPLICABILITY OF BWR/4 TS 3.6.2.5 AND TS 3.6.3.2," USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS (EPID L-2020-LLA-0008) DATED MARCH 2, 2020 DISTRIBUTION:

PUBLIC PM File Copy RidsNrrDorllpl1 Resource RidsNrrDssStsb Resource RidsACRS_MailCTR Resource RidsNrrPMFitzPatrick Resource RidsNrrLALRonewicz Resource RidsRgn1 MailCenter Resource SSmith, NRR ADAMS Access1on No.: ML20024C661 *b>Y memo **b1ye-ma1*1 OFFICE NRR/DORL/LPL 1/PM NRR/DORL/LPL 1/LA NRR/DSS/STSB/BC*

NAME Slee LRonewicz VCusumano (SKrepel for)

DATE 02/27/2020 01/30/2020 01/28/2020 OFFICE OGC-NLO** NRR/DORL/LPL 1/BC NRR/DORL/LPL 1/PM NAME JWachutka JDanna Slee DATE 02/06/2020 03/02/2020 03/02/2020 OFFICIAL RECORD COPY