ML19295G783

Issuance of Amendment No. 331 Regarding Change to Technical Specifications to Remove Ultimate Heat Sink Bar Rack Heaters

ML19295G783
Person / Time
Site:	FitzPatrick
Issue date:	12/19/2019
From:	Samson Lee Plant Licensing Branch 1
To:	Bryan Hanson Exelon Generation Co
Lee S
References
EPID L-2018-LLA-0483
Download: ML19295G783 (17)

Text

UNITED STATES December 19, 2019 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. 331 REGARDING CHANGE TO TECHNICAL SPECIFICATIONS TO REMOVE ULTIMATE HEAT SINK BAR RACK HEATERS (EPID L-2018-LLA-0483)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 331 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 November 29, 2018 (Agencywide Documents Access and Management System No. ML18333A206).

The amendment revises the James A. FitzPatrick Nuclear Power Plant TSs to remove ultimate heat sink bar rack heaters from the operability requirements of the ultimate heat sink.

Specifically, TS Limiting Condition for Operability Condition B requires that all divisions of the required deicing heaters are operable when the ultimate heat sink temperature is

'5._37 degrees Fahrenheit. This amendment removes this requirement from TS Limiting Condition for Operability 3.7.2.

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. 331 to DPR-59

2. Safety Evaluation cc: Listserv

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. 331 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 November 29, 2018, 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. 331, 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 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: December 19, 2019

ATTACHMENT TO LICENSE AMENDMENT NO. 331 JAMES A. FITZPATRICK NUCLEAR POWER PLANT RENEWED FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Replace the following page of the License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

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 3.7.2-1 3.7.2-1 3.7.2-2 3.7.2-2 3.7.2-3 3.7.2-3 3.7.2-4 3.7.2-4

(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. 331, 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 331 Renewed License No. DPR-59

ESW System and UHS 3.7.2 3.7 PLANT SYSTEMS 3.7.2 Emergency Service Water (ESW) System and Ultimate Heat Sink (UHS)

LCO 3.7.2 Two ESW subsystems and UHS shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One ESW subsystem ----------------NOTE-------------------

inoperable. Enter applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources -

Operating," for emergency diesel generator subsystem made inoperable by ESW.

A.1 Restore the ESW 7 days subsystem to OPERABLE status.

(continued)

JAFNPP 3.7.2-1 Amendment 331

ESW System and UHS 3.7.2 ACTIONS ( continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 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. AND B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Both ESW subsystems inoperable for reasons other than Condition A.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the water level in the ESW pump screenwell In accordance with is ~ 236.5 ft mean sea level. the Surveillance Frequency Control Program SR 3.7.2.2 Verify the average water temperature of UHS In accordance with is~ 85°F. the Surveillance Frequency Control Program (continued)

JAFNPP 3.7.2-2 Amendment 331

ESW System and UHS 3.7.2 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.7.2.3 ---------------------------NO TE--------------------------------

1so lati on of flow to individual components does not necessarily render ESW System inoperable.

Verify each ESW subsystem manual, power operated, In accordance with and automatic valve in the flow paths servicing safety the Surveillance related systems or components, that is not locked, Frequency Control sealed, or otherwise secured in position, is in the Program correct position.

(continued)

JAFNPP 3.7.2-3 Amendment 331

ESW System and UHS 3.7.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.2.4 Verify each ESW subsystem actuates on an actual or In accordance with simulated initiation signal. the Surveillance Frequency Control Program JAFNPP 3.7.2-4 Amendment 331

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 331 EXELON FITZPATRICK, LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT TO RENEWED FACILITY OPERATING LICENSE NO. DPR-59

1.0 INTRODUCTION

By letter dated November 29, 2018, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18333A206), Exelon Generation Company, LLC (Exelon, the licensee) submitted a request for changes to the James A. FitzPatrick Nuclear Power Plant (FitzPatrick) Technical Specifications (TSs). The proposed changes would revise the FitzPatrick TSs to remove ultimate heat sink (UHS) bar rack heaters from the operability requirements of the UHS. Specifically, TS 3. 7.2, Limiting Condition for Operability (LCO)

Condition B, requires that all divisions of the required deicing heaters are operable when the UHS temperature is s 37 degrees Fahrenheit (°F). This amendment would remove this requirement from TS LCO 3.7.2.

2.0 REGULATORY EVALUATION

The following explains the use of General Design Criteria (GDC) for FitzPatrick. The construction permit for FitzPatrick was issued by the Atomic Energy Commission (AEC) on May 20, 1970. The original operating license was issued on October 17, 1974, and the renewed operating license was issued September 8, 2008. The plant design criteria for the construction phase is listed in the Updated Final Safety Analysis Report (UFSAR) Chapter 1.5, "Principal Design Criteria." The AEC published the final rule that added Appendix A to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "General Design Criteria for Nuclear Power Plants," in the Federal Register (36 FR 3255) on February 20, 1971, with the rule effective on May 21, 1971. In accordance with a U.S. Nuclear Regulatory Commission (NRC or the Commission) staff requirements memorandum from S. J. Chilk to J.M. Taylor, "SECY-92-223 - Resolution of Deviations Identified During the Systematic Evaluation Program,"

dated September 18, 1992 (ADAMS Accession No. ML003763736), the Commission decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971, which includes FitzPatrick. However, the FitzPatrick UFSAR, Chapter 16.6, "Conformance to AEC Design Criteria," evaluates FitzPatrick against the 10 CFR Part 50, Appendix A GDC.

Enclosure 2

Also, the initial AEC safety evaluation of FitzPatrick, dated November 20, 1972 (ADAMS Accession No. ML19182A200), Chapter 14.0, stated, "Based on our evaluation of the design and design criteria for the James A. FitzPatrick Nuclear Power Plant, we conclude that there is reasonable assurance that the intent of the General Design Criteria for Nuclear Power Plants, published in the Federal Register on May 21, 1971 as Appendix A to 10 CFR Part 50, will be met." Therefore, the NRC staff reviews amendments to the FitzPatrick license using the 10 CFR Part 50, Appendix A GDC unless there are specific criteria identified in the UFSAR.

2.1 Ultimate Heat Sink and Service Water Systems Description The FitzPatrick emergency service water (ESW) and residual heat removal service water (RHRSW) systems are designed to provide cooling water for the removal of heat from equipment served by those systems following a design-basis accident or transient. The ability of the ESW and RHRSW systems to provide adequate cooling for this equipment is an implicit assumption for the safety analyses presented in UFSAR Chapters 5 and 14. The ESW and RHRSW systems, together with the UHS, satisfy Criterion 3 of 10 CFR 50.36( c)(2)(ii).

FitzPatrick draws cooling water from the UHS (Lake Ontario) via a submerged intake structure located about 900 feet out from the lake's shore line. The total design intake flow is 435,000 gallons per minute of which the largest portion is for the main condenser circulating water. The intake is a roofed structure with four segmented shore-facing openings, each 22-feet wide and 8-feet high and each opening with two 11-foot by 11-foot bar rack sections of 11 vertical bars spaced about 10 inches apart, except for the two bar rack sections that were removed in 2017, to block entry of large suspended debris. The top of the intake structure is about 14 feet beneath the lake surface. This submergence avoids the formation of vortices at the lake's surface and minimizes the possibility of floating ice or debris being drawn down from the surface.

The intake bar racks were originally provided with deicing heaters to lessen the likelihood that frazil ice would accumulate on the bar racks and restrict intake flow. Water is drawn through the intake structure and then through the intake tunnel and into the screenwell building. In the screenwell building, the water passes through trash racks that remove the smaller debris that passed through the intake structure inlets and then through traveling screens that prevent small pieces of debris and fish from entering the plant systems. After passing through the trash racks and traveling screens, the water is supplied to the circulating water pumps, ESW pumps, RHRSW pumps, normal service water pumps, the fire protection system, and the makeup demineralizer system.

2.2 Requested Technical Specification Changes The existing FitzPatrick TS 3.7.2 LCO requires two ESW subsystems and UHS to be operable.

TS 3.7.2 Condition B requires two divisions of deicing heaters to be operable while in plant Modes 1, 2, and 3 when the UHS temperature is :S 37 °F. This is intended to ensure that the minimum heat removal capability, assumed in the safety analyses for the system to which the ESW supplies cooling water, will be provided.

TS Surveillance Requirement (SR) 3.7.2.3 requires periodic verification that the required deicing heater electrical current draw is within limits for each division of deicing heaters when the UHS temperature is s 37 °F. TS SR 3. 7.2.5 requires periodic verification that the required deicing heater power is within limits for each division of deicing heaters when the UHS temperature is s 37 °F. These surveillances ensure that adequate heating is being provided at the bar racks to

minimize the accumulation of frazil ice on the bar racks and avoid a potential restriction of water flow. TS SR 3. 7.2.6 requires periodic verification that the required deicing heater resistance to ground is within limits for each division of deicing heaters when the UHS temperature is :s; 37 °F.

This surveillance is performed to monitor long-term degradation of the cable and heater insulation.

With the proposed changes, the existing TS 3.7.2 LCO Condition B will be deleted and Condition C and the associated required actions will be re-lettered from "C" to "B". Existing TS SRs 3.7.2.3, 3.7.2.5, and 3.7.2.6 for the deicing heaters will be deleted and existing TS SRs 3.7.2.4 and 3.7.2.7 will be renumbered accordingly.

2.3 Regulatory Requirements AEC GDC conformance as described in the FitzPatrick UFSAR Section 16.6:

Criteria 44, 45 and 46 - FitzPatrick UFSAR Section 16.6.2.4 and Table 16.6-4, "AEC Design Criteria, Group IV Fluid Systems" describe conformance with Criteria 44, 45, and 46. The Emergency Service Water System is designed to remove heat rejected by the Residual Heat Removal System during plant cooldown, plant shutdown and accident conditions. Equipment cooling water systems are provided to remove heat rejected by the Emergency Diesel Generators and the ECCS. Cooling water requirements for these systems are based on accident cooling water demands of the equipment served. The capability to test the functional performance and to inspect the Emergency Service Water System and vital equipment cooling water systems is provided.

Criterion 2 - The plant equipment that is safety related is designed to permit safe plant operation and to accommodate all design basis accidents without loss of capability to withstand the effects of appropriate natural phenomena at the site.

The regulations under 10 CFR 50.36, "Technical specifications," state that the TSs include items in five specific categories. These categories include, in part: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) SRs; (4) design features; and (5) administrative controls.

Criterion 3 of 10 CFR 50.36(c)(2)(ii) for identifying a TS LCO states:

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.

Regulatory Guide 1.27, "Ultimate Heat Sink for Nuclear Power Plants," regarding ice formation mitigation, states that the potential for adverse environmental conditions, such as icing, should be considered in determining UHS performance.

3.0 TECHNICAL EVALUATION

The license amendment request (LAR) describes the cooling water intake pathway as being through the bar racks of the intake structure, through the intake tunnel into the screenwell building. After entering the screenwell building, water passes through trash racks and traveling

screens, designed to prevent small pieces of debris and fish from entering the cooling water pumps and fouling strainers and heat exchangers. After passing through traveling screens, the water is supplied to the circulating water pumps for main condenser cooling, ESW pumps, RHRSW pumps, normal service water pumps, the fire protection system, and the makeup demineralizer system. The screenwell is configured to allow for "tempering" in winter, where a portion of the warmed discharge water is recirculated back to where the intake water enters the screenwell, upstream of the trash racks and travelling screens. Tempering involves mixing some of the warm discharge water with cold intake water upstream of the trash racks to limit condensate depression or subcooling (how much lower the temperature of condensate leaving the main condenser is than the saturation temperature for the condenser vacuum) related loss of overall plant thermal efficiency and to prevent flow blockage by the buildup of ice on screenwell components. Stray frazil ice that is drawn past the intake to the screenwell will then be melted by the additional heat from the tempering water. The screenwell. can also be configured to allow the flow of circulating water to be reversed in the intake and discharge tunnels to provide a source of suction water for the ESW and RHRSW pumps if the intake pathway should become blocked by large masses of ice.

The LAR describes the icing issue the bar rack heaters were intended to deal with as follows:

The formation of frazil ice on steel bar racks at intake structure openings is common in northern climates. This kind of ice is formed when meteorological conditions are such that the water is supercooled below its freezing point. .. (Supercooling requires a large rate of heat loss, almost always associated with low air temperatures, open water and windy conditions creating lake turbulence and preventing the formation of an ice layer.) Under these conditions, frazil ice can form on intake bar racks, or spongy masses of this ice, formed in other parts of the lake and carried past an intake by wind-driven currents, can adhere to the bar racks. A buildup of frazil ice on intake bars can occur suddenly and can cause a significant reduction in the flow through an intake structure.

During power operation at FitzPatrick under cold weather conditions, the flow of water into the intake may draw in frazil ice. During plant shutdown conditions when the circulating water pumps are not operating, the flow velocity of water into the intake structure is so low that significant frazil ice is not drawn into the intake.

If a rapid buildup of frazil ice on the intake bars were to begin while the circulating water pumps are in operation, the blockage of the intake could potentially challenge the ability of the ESW and RHRSW pumps to perform their safety functions if circulating water pumps were not secured and a manual SCRAM were not inserted in a timely fashion, as required by existing procedures.

Frazil ice will not form on, or adhere to, bar racks that are at a temperature slightly above the freezing point of water. To inhibit the formation of frazil ice, the original plant design included the installation of heating elements in each of the 88 individual bars ... at the intake. Operability of two divisions of the required deicing heaters is required by FitzPatrick Technical Specification (TS) 3.7.2 when the Ultimate Heat Sink (UHS) temperature is s 37 degrees F [°F]. The deicing heater capacity required for compliance with TS 3.7.2 is defined in TS Bases B 3.7.2 as "at least 18 out of the 44 deicing heaters (each heater producing 1670 +/-

10% watts) in each electrical division."

The capacity of the heaters is such as to keep the temperature of the bars at 33 degrees F [°F] during periods when no supercooling occurs or above freezing when up to 1 degree F [°F] supercooling occurs. This is an adequate thermal margin to prevent frazil ice, grease ice, or slush build-up on the bars most of the time.

However, ice buildup can occur on the bars because, under extreme meteorological conditions, the amount of supercooling may exceed the thermal margin available ... There have been five events over the years of plant operation where the buildup of frazil ice on the intake bar racks restricted flow through the intake structure to the point that it caused a plant transient. Two of these events resulted in a lowering of screenwell water level such that a manual trip of the reactor was required. Operator action to downpower the reactor was successful in averting a trip during the other three events.

Following the most recent event (on January 23, 2016), operating procedures were revised to ensure that the potential for frazil ice formation is more closely monitored and that the necessary actions (e.g., tempering) are initiated when appropriate. The recirculation of condenser discharge water through tempering not only maintains the water temperature in the screenwell forebay at a level to eliminate ice in that area but also reduces the flow rate and velocity of the water flowing through the intake, thereby slowing the deposition of frazil ice on plant structures.

The LAR indicates that the licensee had an engineering study performed of the intake system and hazard posed by ice accumulation on the intake bars and what actual benefit the intake bar racks have provided historically and what actions could be taken to optimize reactor safety and plant reliability. The LAR summarizes the study as concluding that debris of dimensions larger than the openings between the bars reaching and becoming an obstruction on the intake structure bar racks has likely not been experienced and that the probability of that ever occurring is low due in part to the position of the intake openings in the water column and intake flow velocity through the bars. Debris smaller than the intake bar rack openings has always been able to enter the intake tunnel and has been removed by the trash racks and traveling screens in the screenwell.

In light of the bar racks likely past and potential future contribution to the intermittent frazil ice caused obstruction of cooling water flow and resulting plant transients, and apparent limited potential benefit, the licensee determined that removal of two of the original eight bar racks (including the deicing heaters for the bars in the removed bar rack sections) would reduce the potential for intake ice blockage caused plant transients while enhancing reliability of the ESW and RHRSW systems. The licensee prepared and implemented a modification (engineering change EC 621217), removing the two bar rack sections. With the two bar rack sections removed, the potential for frazil ice accumulation around those openings is significantly reduced.

The resulting restriction is expected to have a negligible impact for the flow rates drawn by the ESW and RHRSW systems. The LAR indicates that since the deicing heaters are no longer needed to ensure that there is enough flow through the intake for the ESW and RHRSW systems to perform their accident mitigation functions as credited in the safety analyses, there is no need to have a TS LCO or SRs for the deicing heaters in the FitzPatrick TSs.

The intake bar racks are intended to keep large debris from entering the intake tunnel and reaching the screenwell. One strategy employed in other facilities in cold climates with similarly

configured water intakes is removal of bar racks or grating during that portion of the year when frazil ice formation can be expected. That action avoids bars or grating openings being bridged with frazil ice and flow obstructed to the point of affecting facility operation. The FitzPatrick bar heater elements required to be functional were not protective of all bar rack gaps. The design heating capability was enough to mitigate about 1 °F supercooling. The provision of bar rack heaters was intended to avoid the adverse effects of some but not all potential frazil ice formation events by reducing the buildup of frazil ice on the bar racks. With the lake water becoming moderately supercooled by much cooler air blowing along the surface, the frazil ice formation requires nucleation points for ice crystal formation. This can be impurities in the water but is more effectively triggered by tiny ice crystals precipitating out of the moisture in the air due to dropping ambient temperatures and being blown onto the water. As the ice crystals are mixed into the water column and grow, the latent heat of fusion is transferred to the adjacent water and the supercooling diminishes as the bulk water-ice mixture returns to equilibrium and the "stickiness" of the frazil ice is lost. The water-ice mixture passing from the intake structure to the intake tunnel then enters the screenwell as an equilibrium temperature ice-water mixture due to crystal growth and absorbing heat from the intake tunnel that is below the lake bottom.

In the screenwell, the licensee procedurally employs another common strategy for mitigating effects of frazil ice by recirculation of enough warm discharge or tempering water to diminish the subcooling and ice content enough to allow passage through traveling screens without significant pump intake bay level drop. If this is insufficient, a nuclear plant can reduce reactor power and turbine exhaust steam load on the main condenser that allows for a circulating water pump to be stopped. This further reduces the flow of cold water from the UHS resulting in less frazil ice reaching the screenwell and allows more dwell time for the tempering water to better mix with the colder inlet water and frazil ice mixture. The tempering recirculation flow affects frazil ice accumulation on the intake structure inlet openings and bar racks to the extent that flow into the structure is reduced and less active frazil ice is carried to the intake structure inlets, resulting in less obstruction there and less head loss from any obstructing ice.

The staff noted that intake bar rack heaters are not contained in a LCO in any of the Standard Technical Specifications (NUREG-1430, Revision 4, "Standard Technical Specifications -

Babcock and Wilcox Plants"; NUREG-1431, Revision 4, "Standard Technical Specifications -

Westinghouse Plants"; NUREG-1432, Revision 4, "Standard Technical Specifications -

Combustion Engineering Plants"; NUREG-1433, Revision 4, "Standard Technical Specifications

- General Electric Plants (BWR/4)"; and NUREG-1434, Revision 4, "Standard Technical Specifications - General Electric Plants (BWR/6)." The licensee noted in the LAR that only one other nuclear power plant has a TS LCO requirement for intake heaters, Nine Mile Point Nuclear Station, Unit 2, and that there are other plants with intake bar heaters without a TS LCO such as the R. E. Ginna Nuclear Power Plant.

The NRC staff concludes that the intake rack bar heaters do not meet the criterion of 10 CFR 50.36(c)(2) for inclusion in a TS LCO. ESW and RHRSW systems are assumed to be operational 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. The intake rack bar heaters may reduce somewhat the likelihood of frazil ice related blockages of cooling water flow through the intake. However, adequate ESW and RHRSW system flow can be ensured with the removal of the two intake bar rack sections and further by greater attention to cold weather intake water temperature and proactive use of tempering water recirculation as described in the LAR. Based on the rack bar heaters providing negligible contribution to the ESW and RHRSW safety functions, and thus not meeting the criteria of 10 CFR 50.36(c)(2), the staff finds removal of the TS LCO for intake bar rack heaters acceptable.

The regulation under 10 CFR 50.36 states: "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." The licensee may make changes to the TS Bases without prior NRC staff review and approval in accordance with the TS 5.5.11, 'TS Bases Control Program."

Accordingly, along with the proposed TS changes, the licensee also submitted TS Bases changes that are consistent with the proposed TS changes and provide the purpose for each requirement in the TS, consistent with the Commission's Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated July 22, 1993 (58 FR 39132).

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment on September 4, 2019. 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 (January 30, 2019; 84 FR 493). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22( c )(9). Pursuant to 10 CFR 51.22(b ), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

Principal Contributors: Jerome Bettle Matthew Hamm Date: December 19, 2019

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT- ISSUANCE OF

• AMENDMENT NO. 331 REGARDING CHANGE TO TECHNICAL SPECIFICATIONS TO REMOVE ULTIMATE HEAT SINK BAR RACK HEATERS (EPID L-2018-LLA-0483) DATED DECEMBER 19, 2019 DISTRIBUTION:

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