ML20169A510

Issuance of Amendment No. 339 Changes to Technical Specifications Related to Primary Containment Hydrodynamic Loads

ML20169A510
Person / Time
Site:	FitzPatrick
Issue date:	08/20/2020
From:	Justin Poole Plant Licensing Branch 1
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
Samson Lee-NRR/DORL 301-415-3168
References
EPID L-2019-LLA-0197
Download: ML20169A510 (27)

Text

August 20, 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. 339 RE: CHANGES TO THE TECHNICAL SPECIFICATIONS RELATED TO PRIMARY CONTAINMENT HYDRODYNAMIC LOADS (EPID L-2019-LLA-0197)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 339 to Renewed Facility Operating License No. DPR-59 for the James A.

FitzPatrick Nuclear Power Plant. The amendment consists of changes to the technical specifications (TSs) in response to your application dated September 12, 2019, as supplemented by letter dated April 6, 2020.

The amendment makes the following changes: (1) deletes TS 3.6.2.4, Drywell-to-Suppression Chamber Differential Pressure, in its entirety (limiting condition for operation (LCO), associated actions and surveillance requirements); (2) revises TS LCO 3.6.2.2 suppression pool water upper level from 14 feet (ft) to 14.25 ft; and (3) revises the allowable value for TS Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation, Function 3.e., Suppression Pool Water Level - High, from 14.5 ft to 14.75 ft.

A copy of our related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.

Sincerely,

/RA/

Justin C. Poole, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-333

Enclosures:

1. Amendment No. 339 to DPR-59

2. Safety Evaluation cc: Listserv

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. 339 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 September 12, 2019, as supplemented by letter dated April 6, 2020, 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.

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:

2.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 339, 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 120 days.

FOR THE NUCLEAR REGULATORY COMMISSION James G. Danna, Chief Plant Licensing Branch I-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: August 20, 2020 James G.

Danna Digitally signed by James G. Danna Date: 2020.08.20 14:40:46 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 339 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 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.

Remove Pages Insert Pages 3.3.5.1-10 3.3.5.1-10 3.6.2.2-1 3.6.2.2-1 3.6.2.4-1 3.6.2.4-1 3.6.2.4-2 Amendment 339 Renewed License No. DPR-59 (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. 339, 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,

ECCS Instrumentation 3.3.5.1 JAFNPP 3.3.5.1-10 Table 3.3.5.1-1 (page 3 of 5)

Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE 2.

LPCI System (continued) g.

Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass) 1, 2, 3 1 per subsystem E

SR 3.3.5.1.5 SR 3.3.5.1.6 1040 gpm and 1665 gpm h.

Containment Pressure - High 1, 2, 3 4

B SR 3.3.5.1.3 SR 3.3.5.1.6 1 psig and 2.7 psig 3.

High Pressure Coolant Injection (HPCI) System a.

Reactor Vessel Water Level - Low Low (Level 2) 1, 2(c), 3(c) 4 B

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6 126.5 inches b.

Drywell Pressure -

High 1, 2(c), 3(c) 4 B

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6 2.7 psig c.

Reactor Vessel Water Level - High (Level 8) 1, 2(c), 3(c) 2 C

SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6 222.5 inches d.

Condensate Storage Tank Level - Low 1, 2(c), 3(c) 4 D

SR 3.3.5.1.3 SR 3.3.5.1.6 59.5 inches e.

Suppression Pool Water Level - High 1, 2(c), 3(c) 2 D

SR 3.3.5.1.3 SR 3.3.5.1.6 14.75 ft f.

High Pressure Coolant Injection Pump Discharge Flow - Low (Bypass) 1, 2(c), 3(c) 1 E

SR 3.3.5.1.5 SR 3.3.5.1.6 475 gpm and 800 gpm g.

High Pressure Coolant Injection Pump Discharge Pressure - High (Bypass) 1, 2(c), 3(c) 1 E

SR 3.3.5.1.3 SR 3.3.5.1.6 25 psig and 80 psig (continued)

(c)

With reactor steam done pressure > 150 psig.

Amendment 339

Suppression Pool Water Level 3.6.2.2 JAFNPP 3.6.2.2-1 Amendment 3.6 CONTAINMENT SYSTEMS 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression pool water level shall be 13.88 ft and 14.25 ft.

- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

Not required to be met for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during Surveillances that cause suppression pool water level to be outside the limit.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Suppression pool water level not within limits.

A.1 Restore suppression pool water level to within limits.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> B.

Required Action and associated Completion Time not met.

B.1 Be in MODE 3.

AND B.2 Be in MODE 4.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.2.1 Verify suppression pool water level is within limits.

In accordance with the Surveillance Frequency Control Program 339

Drywell-to-Suppression Chamber Differential Pressure 3.6.2.4 JAFNPP 3.6.2.4-1 Amendment 3.6 CONTAINMENT SYSTEMS 3.6.2.4 Deleted 339

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 339 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-59 EXELON FITZPATRICK, LLC EXELON GENERATION COMPANY, LLC JAMES A. FITZPATRICK NUCLEAR POWER PLANT TO RENEWED FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333

1.0 INTRODUCTION

By letter dated September 12, 2019, Agencywide Documents Access and Management System (Reference 1), as supplemented by letter dated April 6, 2020 (Reference 2 and Reference 3)

Exelon Generation Company, LLC (the licensee) submitted a license amendment request (LAR) for the James A. FitzPatrick Nuclear Power Plant (FitzPatrick). The supplement dated April 6, 2020, 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 on December 3, 2019 (84 FR 66231).

The amendment makes the following changes: (1) deletes TS 3.6.2.4, Drywell-to-Suppression Chamber Differential Pressure, in its entirety (limiting condition for operation (LCO), associated actions and surveillance requirements (SRs)); (2) revises TS LCO 3.6.2.2 suppression pool water upper level from 14 feet (ft) to 14.25 ft; and (3) revises the allowable value for TS Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation, Function 3.e., Suppression Pool Water Level - High, from 14.5 ft to 14.75 ft.

2.0 REGULATORY EVALUATION

2.1 Plant Containment Description FitzPatrick is a boiling-water reactor (BWR) plant of the BWR/4 design with a Mark I type pressure suppression primary containment. As described in Section 5.1.2 of the FitzPatrick Final Safety Analysis Report (FSAR), the primary containment consists of the drywell; the pressure suppression pool or torus, which stores a large volume of water; the connecting vent system between the drywell and the torus; isolation valves; the vacuum relief system; and the residual heat removal subsystems for containment cooling. The drywell is a steel pressure vessel in the shape of a lightbulb, and the pressure suppression pool is a torus-shaped steel pressure vessel located below and encircling the drywell. The primary containment houses the reactor vessel, the reactor recirculation system, and other branch connections of the reactor coolant system.

2.2 Primary Containment Hydrodynamic Loads Description The licensee provided the background information about the containment loads in the LAR as follows:

The current Technical Specification requirements for maintenance of a pressure differential between the drywell and torus and the magnitude of the suppression pool water level band were established in response to testing done by General Electric in the early 1970s. These tests identified previously unknown hydrodynamic loads in the torus that would result from a large pipe break in the drywell. The Mark I Torus Program began in 1974 to study these loads and the stresses induced on structures and piping. The Mark I Short Term Program (STP) addressed the initial period of the large break accident, where rapidly expanding steam in the drywell forces air (nitrogen) and steam into the wetwell causing high downward pressure on the torus shell, followed by rapid upward motion of the pool water. This initial break load became known as the pool swell load. Early testing showed that the pool swell load could be mitigated by pressurizing the drywell, which reduced the water volume inside the downcomers, lowering the back pressure resisting the air/steam discharge.

2.3 Requested Technical Specification Changes The proposed TS changes are as follows:

(a) Delete TS 3.6.2.4, Drywell-to-Suppression Chamber Differential Pressure, LCO associated actions and SRs.

(b) Revise TS LCO 3.6.2.2 suppression pool water upper level from 14 ft to 14.25 ft.

(c) Revise the allowable value for TS Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation, Function 3.e., Suppression Pool Water Level - High, from 14.5 ft to 14.75 ft.

2.4 Regulatory Requirements The categories of items required to be in the TSs are provided in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(c). As required by 10 CFR 50.36(c)(2)(i), the TSs will include LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Per 10 CFR 50.36(c)(2)(i), when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.

On July 22, 1993 (58 FR 39132), the Commission published a Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (Final Policy Statement),

which discussed the criteria to determine the items that are required to be included in the TSs as LCOs. The criteria were subsequently incorporated into the regulations by an amendment to 10 CFR 50.36 (60 FR 36953). Specifically, 10 CFR 50.36(c)(2)(ii) requires that a TS LCO be established for each item meeting one or more of the following criteria:

Criterion 1:

Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2:

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.

Criterion 3:

A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4:

A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

The regulations in 10 CFR 50.36(c)(3), require TSs to include items in the category of SRs, which 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 LCOs will be met.

The regulation in 10 CFR 50.36(a)(1) states, in part, A summary statement of the bases or reasons for such specifications shall also be included in the application, but shall not become part of the technical specifications.

Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50 (GDC),

which states, in part:

GDC 4, Environmental and dynamic effects design bases. Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit.

GDC 13, Instrumentation and control. Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems.

Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

GDC 16, Containment design. Reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

GDC 50, Containment design basis. The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident.

Additionally, GDC 38, Containment heat removal, as it relates to the containment heat removal system safety function, which shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss-of-coolant accident (LOCA) and to maintain them at acceptably low levels. The licensee did not include GDC 38 in its LAR; however, the NRC staff determined that this GDC was applicable to the proposed change.

The guidance upon which the NRC staff based its review of the proposed changes is based on NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, Section 6.2.1.1.C, Pressure-Suppression Type BWR Containments, which states the acceptability of pool dynamic loads for plants with Mark I containments is based on conformance with NRC acceptance criteria found in NUREG-0661, Safety Evaluation Report Mark I Containment Long-Term Program (Reference 4).

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the proposed TS changes and their impact on the following aspects of the containment analysis:

short-term pressure and temperature response LOCA hydrodynamic loads safety/relief valve (SRV) loads 3.1 Technical Specification Changes 3.1.1 Deletion of TS 3.6.2.4, Drywell-to-Suppression Chamber Differential Pressure The licensee proposes deletion of the current TS LCO 3.6.2.4, which requires the drywell pressure be maintained 1.7 pounds per square inch (psi) above the pressure of the suppression chamber (torus or wetwell) during Modes 1, 2, and 3. The intent of the proposed change is to remove the drywell-to-suppression chamber differential pressure control during normal plant operation. This change affects the LOCA containment pressure and temperature response, LOCA containment hydrodynamic loads, and safety/relief valve loads.

LCO 3.6.1.4 requires the drywell pressure be maintained 1.95 pounds per square inch gauge (psig) as gauge measurement with reference to the reactor building (secondary containment) pressure during Modes 1, 2, and 3. TS SR 3.6.4.1.1 specifies one of the operability requirements of secondary containment is to verify that it is maintained 0.25 inch of vacuum water gauge measured with reference to the outside atmosphere. Since the outside atmospheric pressure changes, the secondary containment pressure would change. To maintain the drywell and the wetwell pressure within their limits in reference to the drywell pressure, nitrogen must be added or removed from the drywell and the wetwell, which would require opening of the primary containment isolation valves. As discussed in Sections 3.2, 3.3, and 3.4 of this safety evaluation (SE), the NRC staff evaluated the LOCA short-term containment response, LOCA containment hydrodynamic loads, and SRV loads, respectively, and found them acceptable. The staff finds the licensees proposal not to maintain the drywell pressure 1.7 psi above the wetwell pressure acceptable because the licensee has demonstrated through evaluation that it is not limiting for the relevant design-basis events.

The drywell-to-suppression chamber differential pressure does not provide any detection of abnormal degradation of the reactor coolant pressure boundary, and therefore, does not meet 10 CFR 50.36(c)(2)(ii)(A) Criterion 1. The drywell-to-suppression chamber differential pressure does not provide a process variable, design feature, or operating restriction that is an initial condition of a design-basis accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier, and therefore, does not meet 10 CFR 50.36(c)(2)(ii)(B) Criterion 2. Based on the discussion above, the drywell-to-suppression chamber differential pressure is no longer credited as a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The drywell-to-suppression chamber differential pressure, therefore, does not meet 10 CFR 50.36(c)(2)(ii)(C) Criterion 3.

The discussion of Criterion 4 in the Final Policy Statement (58 FR 39137) states that it is the Commissions policy that licensees retain in their TSs, LCOs, action statements, and SRs for the following systems, which operating experience and probabilistic risk assessment have generally shown to be important to public health and safety:

reactor core isolation cooling (RCIC)/isolation condenser residual heat removal standby liquid control recirculation pump trip The drywell-to-suppression chamber differential pressure is not listed in the Final Policy Statement to be important to public health and safety, and recent operating experience or probabilistic risk assessment has not shown this system to be risk-significant. Additionally, the removal of the drywell-to-suppression chamber differential pressure from TSs has no impact on core damage frequency or large early release frequency. As discussed in Section 3.6.1 of this SE, the deletion of the drywell-to-suppression chamber differential pressure has no impact on the offsite and onsite radiological dose consequences of the DBAs, and the requirements of 10 CFR 50.67 continue to be met without the TS drywell-to-suppression chamber differential pressure requirements. Consequently, the drywell-to-suppression chamber differential pressure is not significant to the public health and safety, and therefore, does not meet 10 CFR 50.36(c)(2)(ii)(D) Criterion 4.

Because as discussed above, the drywell-to-suppression chamber differential pressure no longer meets the criteria in 10 CFR 50.36(c)(2)(ii) to establish a TS LCO, the NRC staff finds that it is acceptable to delete TS 3.6.2.4 LCO and its associated action statements and SRs.

The regulation in 10 CFR 50.36(c) states, in part:

Technical specifications will include items in the following categories:

(3)

Surveillance requirements. Surveillance 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.

Since TS 3.6.2.4 LCO no longer meets the requirements in 10 CFR 50.36 and is deleted from the FitzPatrick TSs, the NRC staff finds that it is acceptable to delete its associated SRs, as they no longer support a TS LCO.

3.1.2 Revision in TS LCO 3.6.2.2 Regarding Suppression Pool Water Level In TS LCO 3.6.2.2, the licensee proposes to revise the upper value of the range in which the suppression pool water level must be maintained from 14 ft to 14.25 ft. This change also affects the LOCA containment pressure and temperature response, LOCA containment hydrodynamic loads, and SRV loads.

The proposed change to revise the upper value of the range in which the suppression pool water level must be maintained from 14 ft to 14.25 ft will provide an increased margin for changes in suppression pool water level associated with changes in drywell-to-suppression chamber differential pressure. The increase in the range accommodates suppression pool level change as a result of operations of high-pressure coolant injection and RCIC system testing that adds water to the pool through their turbine exhaust flows. The NRC staff finds the change in the suppression pool level acceptable because the LOCA short-term containment response, containment hydrodynamic loads, and the SRV loads are acceptable, as discussed in Sections 3.2, 3.3, and 3.4 of this SE, respectively.

3.1.3 Revision in TS Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation The licensee proposes to change the allowable value for TS Table 3.3.5.1-1, Function 3.e.,

Suppression Pool Water Level - High, from 14.5 ft to 14.75 ft. The LAR states that the purpose of the change is to maintain a margin consistent with the current margin of 6 inches between the allowable value in TS Table 3.3.5.1-1 and the upper limit of LCO 3.6.2.2 (proposed to increase from 14 ft to 14.25 ft). The LAR states that the Suppression Pool Water Level - High allowable value, 14.75 ft, was proposed in accordance with the licensees setpoint methodology, JAF-CALC-HPCI-00324, 23LS-91A,B Suppression Chamber Level Switch Trip Setpoint, Revision 2, dated February 1999.

The NRC staff reviewed the LAR and the setpoint methodology. The LAR stated that the suppression chamber level switches that perform the function are unchanged; therefore, the associated instrument loop error and the as-found/as-left tolerances will also remain the same.

The NRC staff conducted an independent assessment of the methodology and confirmed the proposed setpoint value with the understanding that the setpoint methodology and uncertainties were not affected by the change. The high-pressure coolant injection high suppression pool water level setpoint shifted up 3 inches from 14.5 ft to 14.75 ft, which is directly correlated to the change in the containment pressure calculation.

The NRC staff finds that the proposed allowable value for TS Table 3.3.5.1-1, Function 3.e.,

Suppression Pool Water Level - High, is acceptable because the approved setpoint methodology was followed, and it supports the associated changes to the suppression pool requirements. The NRC staff also finds this proposed change acceptable to meet GDC 13 because instrumentation is provided to monitor variables and systems over their anticipated ranges for the anticipated operational occurrence (high suppression pool level) and includes appropriate controls to maintain them within prescribed operating ranges.

3.1.4 Technical Specification Bases The regulation in 10 CFR 50.36(a)(1) states, in part, A summary statement of the bases or reasons for such specifications shall also be included in the application, but shall not become part of the technical specifications. Accordingly, along with the proposed TS changes, the licensee also submitted TS Bases changes that correspond to the proposed TS changes for information only. The licensee will make supporting changes to the TS Bases in accordance with TS 5.5.11, Technical Specifications (TS) Bases Control Program.

3.2 Containment Short-Term Analysis for Pressure and Temperature Response During a Design-Basis Accident The containment short-term analysis is performed to determine the containment peak pressure and temperature during the most limiting DBA. As the licensee stated in Reference 2, all initial conditions, key inputs, and assumptions used in the design-basis LOCA short-term containment pressure and temperature response can be found in Sections 2.1 and 2.2 as found in Reference 3 (Reference 3 is proprietary, non-public; Reference 2 is the nonproprietary, public version of this document).

The short-term analysis covers the LOCA blowdown period during which the maximum drywell pressure and maximum drywell-to-suppression chamber differential pressure occur. The licensee used the analysis of record (AOR) documented in NEDC-33087P, Revision 1 (Reference 5) methods (i.e., LAMB code (Reference 6)) for the LOCA containment mass and energy release, and the M3CPT code (Reference 7) for containment response. The NRC accepted the AOR methods for the short-term analysis in FitzPatrick License Amendment No. 287 issued May 17, 2007 (Reference 8).

The analysis initial conditions listed in Table 1 of Attachment 2 to the licensees letter dated April 6, 2020 (Reference 3), are conservatively selected. The values of the key initial conditions are (a) thermal power of 2,587 megawatts thermal (MWt) (102 percent of the rated thermal power); (b) maximum reactor dome pressure of 1,060 pounds per square inch absolute (psia) during normal operation; (c) drywell relative humidity of 20 percent; (d) suppression pool volume, which includes the water volume in downcomers; (e) suppression pool level 14.25 ft; and (f) subcooled break fluid at pressure and temperature of 1,060 psia and 423.9 degrees Fahrenheit (°F), respectively, for mass and energy release, which is conservative for maximizing the mass released because of its higher density than saturated fluid.

The AOR documented in NEDC-33087P, Revision 1, Section 9.0 (Reference 5), provides evaluation and results of the following short-term analysis cases of design-basis LOCA containment pressure and temperature response for normal feedwater temperature and final feedwater temperature reduction. The results of these analyses were used for evaluating the containment hydrodynamic loads.

Case 1, which corresponds to 102 percent of current licensed thermal power (CLTP) and 100 percent of rated core flow (RCF).

Case 2, which corresponds to 102 percent of CLTP with 105 percent of RCF (i.e.,

increased core flow).

Case 3, which corresponds to 102 percent of CLTP with 79.8 percent of RCF (i.e., on the maximum extended load line limit analysis (MELLLA) line).

Case 4, which corresponds to 62 percent of CLTP with 36.8 percent of RCF (minimum pump speed on the MELLLA line).

In Reference 2, the licensee stated that Cases 1 and 2 are the bounding cases for this analysis.

Containment peak pressure and temperature were recalculated for both cases, and the results are provided in Table 3 of Reference 3. The bounding pressure and temperature histories are plotted in Figures 11 and 12 of Reference 3.

Based on the proposed changes, the licensee reanalyzed the containment peak pressure and temperature for Cases 1 and 2 only because Case 2 is the AOR most limiting case, and the Case 1 results are very close to the Case 2 results in the AOR, using the AOR methods described above, with conservative initial conditions and key inputs listed in Table 1 of General Electric Hitachi (GEH) Technical Report 005N1724-P, Revision 1 (Reference 3). Table 3 of GEH 005N1724-P, Revision 1 (Reference 3), tabulates the peak drywell pressure and temperature results. The NRC staff reviewed Table 3 of Reference 3, which shows small increases from AOR. The maximum peak drywell pressure increased from 54.5 psia to 57.8 psia, and the peak drywell temperature increased from 285.9 °F to 289.6 °F, which are both at 102 percent CLTP and 105 percent RCF. The results show that the new peak values of the containment pressure and temperature are within the design limits of 56 psig (70.7 psia) and 309.6 °F, respectively, given in the FitzPatrick FSAR, Table 5.2-1.

The NRC staff finds the revised short-term analysis with the proposed changes acceptable because by using NRC-accepted methods, the licensees calculated peak pressure and temperature are within the containment design pressure and temperature limits.

3.3 Loss-of-Coolant Accident Containment Hydrodynamic Loads The AOR documented in NEDC-33087P, Revision 1 (Reference 5), Section 9.0, provides assessment of the following containment LOCA hydrodynamic loads based on the short-term containment pressure and temperature response analysis:

vent system thrust loads pool swell loads condensation oscillation loads chugging loads For the proposed changes, the NRC staffs evaluation of these loads is given below.

3.3.1 Vent System Thrust Loads During a design-basis LOCA, due to a rapid increase in the drywell pressure, the water initially in the vent system is accelerated into the suppression pool and clears the vent system. The pressure difference between the vent system and the wetwell during the vent clearing process causes structural loads on the vent system. The licensee evaluated the vent system thrust loads in Section 4.4 of Reference 3.

For the proposed changes, the licensee recalculated the vent system thrust loads using the key inputs and initial conditions given in Table 1 of GEH Report 005N1724-P, Revision 1 (Reference 3), for the limiting cases. These are Cases 1 and 2 as described in Section 3.2 of this SE, as indicated by the licensee (Reference 2). Comparison of the calculated maximum loads listed in Table 2 of GEH Report 005N1724-P, Revision 1 (Reference 3), to the current design and licensing basis loads shows that the recalculated vent thrust forces F3V, F3VT, and F3H for the downcomer mitre bend exceed the current design and licensing basis loads in the AOR. The remaining loads, as shown in Table 2 of GEH Report 005N1724-P, Revision 1 (Reference 3), are bounded by the current design and licensing basis loads in the AOR. The vent system thrust load condition adjustment factor is listed in Tables 13 and 14 of the Imperia Report (Attachment 6 to Reference 1).

The NRC staff finds the revised vent system thrust loads analysis with the proposed changes acceptable because by using NRC-accepted methods, the licensees calculated maximum vent thrust loads are within the current design and licensing basis limits, except for the downcomer mitre bend forces F3V, F3VT, and F3H, which exceed the current licensing basis loads. The NRC staff evaluated the load exceedance in Section 3.8.3 of this SE and found it acceptable.

3.3.2 Pool Swell Loads The pool swell loads depend on the maximum pressurization rate of a design-basis LOCA. The licensee evaluated the pool swell loads in Section 4.1 of Reference 3. Based on the proposed conservative assumption of 0.0 psi initial operating drywell-to-suppression chamber differential pressure, the licensee calculated a maximum drywell pressurization rate and compared that with the test data in NEDO-21944 (Reference 10). The licensee found that the calculated value was bounded by the test data, which is part of the existing licensee analysis. Therefore, the licensee confirmed that the existing pool swell test data remains applicable. The NRC staff compared the calculated value with the test data and agrees that the calculated value was bounded by the test data.

NEDO-21888, Revision 2 (Reference 11), Section 4.3, describes the following primary loads associated with the pool swell phenomena:

torus net vertical loads torus shell pressures vent system impact and drag loads impact and drag loads on other structures above the pool froth impingement loads pool fallback load downcomer water clearing jet load on submerged structures LOCA bubble-induced drag loads on submerged structures The load condition adjustment factors for the above loads are listed in Tables 13 and 14 of the Imperia Report (Attachment 6 to Reference 1).

The NRC staff finds that the licensee appropriately determined the pool swell load condition adjustment factors as stated in Section 4.3.1.1 of NEDO-21888, Revision 2.

3.3.3 Condensation Oscillation Loads Following the pool swell transient during a design-basis LOCA, there is a period during which condensation oscillation occurs in the downcomer exit. The variation in the steam condensation rate causes an oscillatory motion of the steam-water interface, which results in pressure oscillation imposing pressure loads on the torus shell, vent system, and submerged structures.

The maximum loads are evaluated from the maximum of the root mean square (RMS) value of the oscillating pressure. The licensee evaluated the condensation oscillation loads in Section 4.2 of Reference 3.

For the proposed change, the licensee calculated the maximum of the RMS value of the pressure oscillation and determined that the calculated value was bounded by the full-scale test data in NEDO-24539 (Reference 9), which is part of the existing licensee analysis. The NRC staff compared the calculated value with the test data and agrees that the test data bounded the calculated value. Therefore, the AOR condensation oscillation loads remain valid.

The NRC staff finds it acceptable that the AOR condensation oscillation pressure loads on the torus shell, vent system, and submerged structures remain valid because the condensation oscillation maximum RMS pressure test value bounds the calculated value with the proposed changes.

3.3.4 Chugging Loads During a design-basis LOCA, the chugging phenomena refers to the unsteady condensation process, which occurs late in the blowdown when the vent system flow rates become small.

Chugging takes place when the steam flow rate through the vent system falls below a rate necessary to maintain steady condensation at the downcomer exits. A higher condensation rate causes the pool water to re-enter the downcomers. During chugging, steam bubbles form at the downcomer exits, oscillate as they grow, and then begin to collapse, which results in a variable frequency oscillating load on the torus shell. The licensee evaluated the chugging loads in Section 4.3 of Reference 3.

The AOR chugging loads are derived from the full-scale test facility test results (Reference 9),

and the licensee determined that the AOR chugging loads remain valid. The staff agrees with the licensees assessment because it is conservative and the loads are based on test results, which are part of the existing licensee analysis. The NRC staff finds the chugging loads in the AOR remain valid for the proposed changes.

3.4 Safety/Relief Valve Containment Loads Actuation of an SRV causes pressure and thrust loads on the SRV discharge line and the T-quencher in the suppression pool. In addition, water and then discharge of nitrogen into the suppression pool through the T-quencher would result in pressure loads on the submerged portion of the torus shell, and drag and jet loads on the submerged structures.

The SRV loads are based on (a) first SRV actuations followed by (b) subsequent SRV actuations in some cases. For both initial and subsequent actuations, the loads can be divided into two categories. The first category includes the internal pressure loads and thrust loads on the SRV discharge line and T-quencher. The second category consists of loads resulting from air-bubble formation in the suppression pool following water and air clearing. The air-bubble exerts pressure load on the pool boundaries and drag load on the submerged structures.

3.4.1 Safety Relief Valve Discharge Line Loads As stated in Section 2.2 of the Imperia Report (Attachment 6 to Reference 1), the AOR bounding worst case SRV load condition is the small break accident/intermediate break accident first actuation combined with a small break accident/intermediate break accident reflood case when the steam is in the drywell, which produces the highest reflood height because of steam (instead of air) entering the SRV discharge line through the vacuum breakers after the first actuation. Section 3.2 of the Imperia Report (Attachment 6 to Reference 1) states that the reflood height in the AOR analysis is significantly greater than the reflood height, based on the proposed change. Therefore, the NRC staff finds that the AOR bounding case load condition for SRV discharge line loads remains valid.

3.4.2 Safety Relief Valve Bubble Drag Loads For the effect of the proposed changes on the SRV bubble drag loads, Section 5.2.1.1.c of NEDO-21888, Revision 2 (Reference 11), states that the bubble drag load is a function of pressure at the air/water interface, which is a function of the main steam line pressure. In addition, the bubble drag load on the torus and submerged structures is a function of distance from the T-quencher and is independent of submergence. Therefore, the NRC staff finds that the SRV bubble drag loads are not affected because these loads do not depend on the proposed changes.

3.4.3 Safety Relief Valve Jet Loads For the SRV jet loads, the jet velocity at the submerged structure depends on the distance, and it decreases rapidly with the distance from the location of its discharge. The bubble drag load depends on the stagnation pressure and the projected area of the submerged structure. The stagnation pressure depends on the square of the jet velocity in the vicinity of the structure. The licensee stated in the Imperia Report (Attachment 6 to Reference 1) that per the FitzPatrick unique loads analysis report, the jet loads are significantly smaller and are, therefore, not considered, as they are bounded by the bubble drag loads. The NRC staff agrees that the jet loads are bounded by the bubble drag load, and therefore, do not need to be considered.

3.4.4 Safety Relief Valve Load Condition Adjustment Factors The SRV load condition adjustment factors are given in Tables 13 and 14 of the Imperia Report (Attachment 6 to Reference 1). The NRC staff reviewed Tables 13 and 14 and found that the licensee appropriately determined the SRV load condition adjustment factors and applied them to the AOR loads.

3.5 Main Steam Line Break Response The licensee evaluated the largest main steam line break. As stated in NUREG-0661 (Reference 4), the DBA for Mark I containment designs is the instantaneous guillotine rupture of the largest pipe in the reactor coolant system, which is the recirculation suction line break (RSLB) LOCA. As discussed in Section 2.1 of Reference 3, the licensee re-ran the analysis, assuming the largest main steam line break to confirm that the RSLB LOCA is limiting. The NRC staff agrees with the licensees confirmatory analysis as described in Reference 3 because it is a reasonable approach to evaluate limiting conditions. The licensees result is also consistent with NUREG-0661.

The NRC staff finds it acceptable that the design-basis containment hydrodynamic loads for the proposed changes be based on RSLB LOCA.

3.6 Environmental Qualification The environmental qualification (EQ) parameters that determine the EQ profile are pressure, temperature, relative humidity, and radiological dose. The licensee stated the following:

(a) current EQ accident pressure envelope continues to bound the revised peak short-term accident pressure of 57.8 psia; (b) the EQ accident temperature envelope is not significantly affected by the change in normal operating parameters associated with this change; and (c) the normal, abnormal, and accident relative humidity is assumed to be 100 percent in the suppression pool when the plant is operating, and therefore, will not be impacted.

The NRC staff finds the proposed changes acceptable because the proposed changes do not affect the currently established EQ profile for pressure, temperature, and relative humidity because the normal, abnormal, and accident service conditions are not impacted or remain bounded.

3.6.1 Radiological Dose Analyses The deletion of the drywell-to-suppression chamber differential pressure has no impact on the offsite and onsite radiological dose consequences of the DBAs, and the requirements of 10 CFR 50.67 continue to be met without the TS drywell-to-suppression chamber differential pressure requirements. Consequently, the drywell-to-suppression chamber differential pressure is not significant to the public health and safety, and therefore, does not meet 10 CFR 50.36(c)(2)(ii) Criterion 4. Changes approved in this license amendment will be bounded by the current design-basis analyses and will have either negligible or no impact on the radiological consequences associated with the FitzPatrick DBAs.

3.7 Net Positive Suction Head The net positive suction head (NPSH) from the pumps that draw water from the suppression pool during an accident would be affected by a change in the suppression pool level and its peak temperature during the accident. The pumps that draw water from the suppression pool during accidents are the emergency core cooling system (ECCS) and RCIC system pumps.

The licensee stated that the AOR for these pumps uses the minimum initial suppression pool water level of 13.88 ft. The peak temperature of the suppression pool occurs during the long-term peak suppression for which the AOR is not affected.

The NRC staff finds the NPSH evaluation for the ECCS and RCIC pumps acceptable because the AOR for the minimum suppression pool level and the long-term peak suppression pool temperature is not affected based on the proposed TS changes.

3.8 Structural Evaluation 3.8.1 Drywell From LAR Attachment 1, Section 3.1, the NRC staff noted that the licensee reevaluated GEH Technical Report 005N1724, Revision 0, for the short-term containment response to the limiting design-basis LOCA for zero drywell-to-torus differential pressure and a suppression pool high water level of 14.25 ft (3-inch increase). This reevaluation used the same NRC-approved methodology in the previous AOR and determined that new values of peak pressure (57.8 psia) and peak temperature (289.6 °F) remain bounded by the drywell design values (70.7 psia (56 psig) and 309 °F). The staff finds that there is reasonable assurance that drywell structural integrity will be maintained as a result of the requested TS changes because the reevaluated values of peak pressure and temperature remain bounded by the corresponding drywell design values in the current licensing basis (FitzPatrick FSAR, Table 5.2-1).

3.8.2 Suppression Pool (Torus) Structural Elements and Structural Element Supports The NRC staff reviewed LAR Attachment 1, Section 3.2, Structural Evaluation, and the supporting Technical Report 13-0541-TR-002 included as LAR Attachment 6. The staff noted this technical report documented a detailed explanation of inputs, methodologies, and calculations to demonstrate, consistent with applicable code requirements, the structural adequacy of the torus structural elements for the LAR-proposed new normal operating parameters (i.e., zero differential pressure and 3-inch increase in torus high water level to 14.25 ft). The staff also noted that the methodology of this report used the current Plant Unique Analysis Report or PUAR (TR-5321-1 and TR-5321-2 (LAR Attachment 6, References 7.3.2 and 7.3.3)) from the FitzPatrick Mark I Containment Program (based on 1.7 psi drywell-torus differential pressure) and applied appropriate adjustment factors to account for zero differential pressure, the 3-inch torus water level increase, and other applicable plant changes (power uprate, water in vent pipe bowl, torus new corrosion allowance of 0.143 inch, upgrade of torus ECCS and RCIC suction strainers, and high-pressure coolant injection steam discharge sparger modification) since the end of the Mark 1 program. The staff further noted from LAR Attachment 1, Section 3.1, that containment hydrodynamic loads (pool swell, chugging, condensation oscillation, and vent system thrust) were reevaluated in GEH Technical Report 005N1724 using NRC-approved methodology in the AOR for the new operating parameters and provided input for the structural evaluations of Technical Report 13-0541-TR-002.

Load Condition Adjustments The final load condition adjustments for new operating parameters and considering plant modifications are described in detail in LAR Attachment 6, Section 2.10, and summarized in Table 13, Load Condition Adjustment Factors for Structural Element Evaluation, and Table 14, Load Condition Adjustment Factors for +3-inches Increased Submergence, of LAR. Bounding event combination results consistent with the FitzPatrick PUAR were adjusted based on these factors, and resulting stresses were compared to the corresponding acceptance criteria (e.g., allowable stresses) in the construction codes of record discussed below.

Construction Codes of Record The staff noted from Section 2.12, Updated Construction Code, of LAR Attachment 6 that the construction codes of record, as reconciled by the FitzPatrick Mark 1 long-term-program and consistent with the structural acceptance criteria in the NRC SE in NUREG-0661 for the Mark 1 program, are as follows:

Primary containment pressure boundary (drywell and torus) structural elements: ASME Boiler and Pressure Vessel Code (ASME Code),Section III, Division 1, 1977 Edition.

(This construction code of record is changed to a later code edition, namely the 2007 Edition with 2008 Addenda in this LAR, as discussed further below).

Structural element supports: American Institute of Steel Construction (AISC), Manual of Steel Construction, - 7th Edition, 1970. (There is no change made to this construction code of record by this LAR.)

For evaluation of the TS changes in this LAR, the licensee changed the construction code of record for the primary containment structural elements to the ASME Code,Section III, Division 1, Subsection NE, 2007 Edition with 2008 Addenda, which is incorporated by reference in 10 CFR 50.55a. The licensee stated that although there are no changes in the code equations for evaluation of the zero differential pressure between the 1977 Edition and the 2007 Edition with 2008 Addenda of the ASME Code, the advantage of the later code edition is that the material properties are consistent, and the allowable stress intensity values are approximately 14 percent higher in the later code edition (based on a safety factor of 3.5 on the ultimate tensile stress (Su) versus a safety factor of 4.0 in the 1977 Edition).

The staff evaluated the change in edition of the primary containment code of record and finds the use of the 2007 Edition with 2008 Addenda of the ASME Code,Section III, Division 1, for structural evaluation of the primary containment structural elements acceptable because it is incorporated by reference in 10 CFR 50.55a(a), and therefore, is approved by the NRC.

Structural Element Evaluation Results The structural elements evaluated by the licensee are consistent with the FitzPatrick Mark 1 Containment Program and consist of the torus shell, torus support structures and attachment welds, internal ring girder, vent system (vent pipe, vent header, downcomers, vent header deflector), T-quencher, and other miscellaneous internal structures (i.e., catwalk, spray header, vent pipe bellows, and monorail).

The staff reviewed the licensees evaluation results of the torus structural elements and supports to code requirements in Sections 5.3 through 5.7 and Section 5.9, Miscellaneous Structures, of LAR Attachment 6. The staff noted that Table 38 of LAR Attachment 6 summarizes the code allowable to actual ratios for controlling event combinations from its evaluation of the structural elements for the new zero pressure differential normal operating condition. The staff noted that for the structural elements, the controlling stress ratios for the torus pressure boundary and support structural elements are 1.14 for the torus shell and 1.13 for support elements, and the range of all stress ratios varies from 1.06 (monorail) to 2.73 (main vent/drywell intersection) and 4.73 for vent pipe bellows displacement. The licensee concluded that the summary of results demonstrates the acceptability of all structural elements for continued service of the FitzPatrick primary containment at zero drywell-torus differential pressure during normal operation by meeting code requirements with margin. The staff further noted that based on the available margin in Table 38 of LAR Attachment 6 and the maximum load condition adjustment factor of 1.056 for +3-inch increased submergence in Table 14 of LAR Attachment 6, the licensee concluded that code requirements are met for both the zero drywell-torus pressure differential and the 3-inch increase in torus water level conditions sought in the LAR.

The NRC staff finds that there is reasonable assurance that the structural and functional capability of the primary containment (drywell, torus structural elements, and torus supports) to withstand the reanalyzed hydrodynamic loading conditions will be maintained for zero drywell-torus differential pressure and the 3-inch increase in torus water level during normal operation, because the licensees reevaluation of the changes consistent with the methodology in the current licensing basis demonstrated that the primary containment structural elements continue to meet acceptance criteria in the construction codes of record, as updated, with margin. Therefore, based on its review, the staff concludes that there is reasonable assurance that the regulatory requirements in GDC 4, 16, and 50 will continue to be met for the FitzPatrick primary containment with the new operating parameters.

3.8.3 Torus Attached Piping Section 3.0 of Attachment 2 to the licensees April 6, 2020, letter (Reference 3) discussed the evaluation method that the licensee has used previously in the AOR and has been approved by the NRC. The results of the containment hydrodynamic loads are discussed in Attachment 2, Section 4, of Reference 3. Condensation oscillation loads are evaluated by comparing the root mean square of the pressure oscillations to the full-scale test data. This comparison shows that the existing condensation oscillation loads remain valid. Chugging loads are derived from the full-scale test facility test results. The test conditions bound the conditions in Table 1 of of Reference 3. The existing chugging loads remain valid. The vent thrust loads were recalculated to compare with the current licensing basis loads.

The comparisons in Table 2 of Attachment 2 of Reference 3 show that the vent thrust loads for the downcomer mitre bend exceed by less than 10 percent of the current design and licensing basis loads. As discussed in Section 3.8.2 of this SE, the licensee takes advantage of the later code edition to increase the allowable stress intensity values to be 14 percent higher than the allowable stress intensity of the original code edition. Therefore, the NRC staff concludes that the 10 percent load exceedance is acceptable.

As shown in Table 2 of Attachment 2 of Reference 3, all other loads are still less than the loads in the plant unique load definition report for operating delta P. Pool swell loads are evaluated by comparing the pressurization rate during the vent clearing phase to the pressurization rate in the quarter scale suppression pool swell test data. The maximum pressurization rate is calculated for zero initial drywell-to-suppression chamber pressure differential. This is below the pressurization rate derived from the test data, which confirms that the existing pool swell test data is still valid for FitzPatrick. The suppression pool loads are calculated by applying an appropriate factor that was previously approved by NRC.

The staff reviewed Technical Report 13-0541-TR-002 (LAR Attachment 6), which demonstrates the structural adequacy of the torus structural elements and torus attached piping for the new, proposed normal operating parameters, with specific focus on Sections 3.3, 5.8, and 5.10 related to evaluation of torus-attached piping and its supports. The methodology of this report uses the current Mark I Containment Program analysis and applies appropriate factors to account for 0 psid delta P, the torus water level increase, and all other applicable plant changes since the end of the Mark 1 program. The staff reviewed the topical report detailed explanation of inputs, methodologies, calculations, and conclusions. The staff finds that all applicable structural elements and torus attached piping continue to meet code requirements with adequate margin under the new, normal operating parameters.

3.9 NRC Staff Conclusion

of Proposed Changes The NRC staff reviewed the proposed changes, which are the deletion of TS LCO 3.6.2.4 and a change in TS LCO 3.6.2.2 upper suppression pool level from 14 ft to 14.25 ft, and finds them acceptable. The staff also reviewed the revision in TS Table 3.3.5.1-1, Function 3.e.,

Suppression Pool Water Level - High, increased from 14.5 ft to 14.75 ft, and finds it acceptable.

The NRC staff reviewed the impacts of the proposed TS changes on (a) short-term containment pressure and temperature response, and (b) containment hydrodynamic loads. The conclusions are as follows:

The revised short-term pressure and temperature response is acceptable because the peak pressure and temperature remain bounded by their design limits.

The AOR pool swell, condensation oscillation, and chugging loads remain valid.

The vent system thrust loads remain within the current licensing basis limits except for the downcomer mitre bend forces F3V, F3VT, and F3H, which are exceeded but acceptable.

The AOR SRV loads remain valid.

The current EQ profile is not affected and remains valid.

The AOR NPSH for the pumps that draw water from the suppression pool is not affected.

The NRC staff concludes that with the deletion of TS LCO 3.6.2.4, FitzPatrick continues to meet the 10 CFR 50.36 requirements.

The NRC staff concludes that FitzPatrick continues to meet the requirements of GDC 4, 13, 16, 38, and 50 as follows:

GDC 4 - because the SRV loads and LOCA containment hydrodynamic loads are acceptable.

GDC 13 - because instrumentation is provided to monitor variables and systems over their anticipated ranges for the anticipated operational occurrence (high suppression pool level) and includes appropriate controls to maintain them within prescribed operating ranges.

GDC 16 - because the containment design conditions important to safety are not exceeded, as demonstrated by the acceptable LOCA short-term containment pressure and temperature response.

GDC 38 - because the containment heat removal system is not affected since there is no impact on the NPSH of the ECCS pumps that draw water from the suppression pool during a LOCA or abnormal events.

GDC 50 - because the design leakage rate will not be exceeded since the LOCA short-term containment response is acceptable, and the containment heat removal system is not affected.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the New York State official was notified of the proposed issuance of the amendment on April 14, 2020. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

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 SRs.

The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (December 3, 2019; 84 FR 66231). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

7.0 REFERENCES

1. Letter from Exelon Generation to U.S. NRC, License Amendment Request - Proposed Changes to the Technical Specifications Related to Primary Containment Hydrodynamic Loads, dated September 12, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19255D988).

2. Letter from Exelon Generation to U.S. NRC, Response to Request for Additional Information in Support of License Amendment Request - Proposed Changes to the Technical Specifications to Primary Containment Hydrodynamic Loads, dated April 6, 2020 (ADAMS Accession No. ML20098D248).

3. General Electric Hitachi (GEH) 005N1724-P, Revision 1, Proprietary Report, Exelon Generation Company LLC, James A. FitzPatrick Nuclear Power Plant Short-Term Containment Analysis for Zero Drywell-to-Wetwell Pressure Differential, dated March 2020 (ADAMS Accession No. ML20098D250) (non-public, proprietary).

4. 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).

5. General Electric NEDC-33087P, Revision 1, J. A. Fitzpatrick Nuclear Power Plant, APRM/RBM/Technical Specifications/Maximum Extended Operating Domain (ARTS/MEOD), dated September 2005 (ADAMS Accession Nos. ML060390374 (non-public, proprietary) and ML060390372 (public, nonproprietary)).

6. General Electric Company NEDE-20566-P-A, Class III, General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50 Appendix K, dated September 1986 (ADAMS Accession No. ML092110816 (non-public, proprietary).

7. General Electric, Licensing Topical Report, The General Electric Mark III Pressure Suppression Containment Analytical Model, NEDO-20533, dated June 1974, and Supplement 1, dated September 1975 (ADAMS Legacy Library Accession Nos. 8707090439 and 8707110040).

8. Letter from NRC to Entergy Nuclear Operations, Inc., James A. FitzPatrick Nuclear Power Plant - Issuance of Amendment Re: Implementation of the Average Power Range Monitor, Rod Block Monitor Technical Specification Improvements with the Maximum Extended Operating Domain Analysis (TAC No. MC9681), dated May 17, 2007 (ADAMS Accession No. ML070430065).

9. General Electric NEDO-24539, 79NED84, Class I, Mark I Containment Program, Full Scale Test Program Final Report, Task Number 5.11, dated August 1979 (ADAMS Accession No. ML19254C695.

10. General Electric NEDO-21944, Mark I Containment Program Quarter Scale Plant Unique Tests, dated June 1979 (ADAMS Package Accession No. ML19254F709.

11. General Electric, NEDO-21888, 81NED282, Class I, Revision 2, Mark I Containment Program Load Definition Report, dated November 1981 (ADAMS Accession No. ML20141L528).

Principal Contributors: A. Sallman K. West G. Thomas K. Hsu K. Bucholtz S. Meighan Date: August 20, 2020

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