Text

Application to Revise Technical Specification Limiting Condition of Operation 3.5.2, ECCS – Operating, Note 1 to Include Residual Heat Removal Pump Flow Paths (SQN-TS-23-04 and WBN-TS-23-020)
	ML24317A243
Person / Time
Site:	Sequoyah, Watts Bar
Issue date:	11/12/2024
From:	Hulvey K; Tennessee Valley Authority
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	CNL-24-021
	Download: ML24317A243 (1)
	v • d • e

10 CFR 50.90 1101 Market Street, Chattanooga, Tennessee 37402 CNL-24-021 November 12, 2024 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327 and 50-328 Watts Bar Nuclear Plant, Units 1 and 2 Facility Operating License Nos. NPF-90 and NPF-96 NRC Docket Nos. 50-390 and 50-391

Subject:

Application to Revise Technical Specification Limiting Condition of Operation 3.5.2, ECCS - Operating, Note 1 to Include Residual Heat Removal Pump Flow Paths for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-04 and WBN-TS-23-020)

In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, Application for amendment of license, construction permit, or early site permit, Tennessee Valley Authority (TVA) is submitting a request for an amendment to Renewed Facility Operating License Nos. DPR-77 and DPR-79 for the Sequoyah Nuclear Plant (SQN), Units 1 and 2; and Facility Operating License Nos. NPF-90 and NPF-96 for the Watts Bar Nuclear Plant (WBN), Units 1 and 2, respectively.

The proposed license amendment would revise SQN Units 1 and 2 and WBN Units 1 and 2 Technical Specification (TS) Limiting Condition of Operation (LCO) 3.5.2, ECCS -

Operating, Note 1 to include residual heat removal (RHR) pump flow paths as follows.

In MODE 3, both safety injection (SI) pump flow paths and one residual heat removal (RHR) pump flow path may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform pressure isolation valve testing per SR 3.4.14.1.

The enclosure to this submittal provides a description and assessment of the proposed change, a regulatory evaluation, and a discussion of environmental considerations. provides a marked-up version of the affected TS pages of SQN Units 1 and 2 showing the proposed changes. Attachment 2 provides a marked-up version of the affected

U.S. Nuclear Regulatory Commission CNL-24-021 Page 2 November 12, 2024 TS pages of WBN Units 1 and 2 showing the proposed changes. Attachment 3 provides a marked-up version of the SQN Units 1 and 2 TS Bases showing the proposed changes. provides a marked-up version of the WBN Units 1 and 2 TS Bases showing the proposed changes. Changes to the existing TS Bases are provided for information only and will be implemented under the Technical Specification Bases Control Program.

TVA requests approval of the proposed license amendment within one year from the date of this submittal with implementation within 60 days of issuance of the amendment.

TVA has determined that there are no significant hazards considerations associated with the proposed change and that the TS change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). In accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and enclosure to the Tennessee State Department of Environment and Conservation.

There are no new regulatory commitments contained in this letter. Please address any questions regarding this request to Amber V. Aboulfaida, Senior Manager, Fleet Licensing, at avaboulfaida@tva.gov.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 12th day of November 2024.

Respectfully, Kimberly D. Hulvey General Manager, Nuclear Regulatory Affairs & Emergency Preparedness

Enclosure:

Description and Assessment of the Proposed Change cc (Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Sequoyah Nuclear Plant NRC Senior Resident Inspector - Watts Bar Nuclear Plant NRC Project Manager - Sequoyah Nuclear Plant NRC Project Manager - Watts Bar Nuclear Plant Director, Division of Radiological Health - Tennessee Department of Environment and Conservation Digitally signed by Edmondson, Carla Date: 2024.11.12 10:21:26

-05'00'

CNL-24-021 Enclosure Description and Assessment of the Proposed Change

Subject:

Application to Revise Technical Specification Limiting Condition of Operation 3.5.2 ECCS - Operating, Note 1 to include Residual Heat Removal Pump Flow Paths for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-04 and WBN-TS-23-020)

CONTENTS 1.0

SUMMARY

DESCRIPTION............................................................................................... 1 2.0 DETAILED DESCRIPTION................................................................................................ 1 2.1 System Design and Operation........................................................................................ 1 2.2 Reason for the Proposed Change.................................................................................. 2 2.3 Description of the Proposed Change.............................................................................. 4

3.0 TECHNICAL EVALUATION

............................................................................................... 5

4.0 REGULATORY EVALUATION

........................................................................................... 7 4.1 Applicable Regulatory Requirements and Criteria.......................................................... 7 4.2 Precedent....................................................................................................................... 8 4.3 No Significant Hazards Consideration Determination Analysis...................................... 9 4.4 Conclusion.................................................................................................................... 11

5.0 ENVIRONMENTAL CONSIDERATION

........................................................................... 11 Attachments 1.

Proposed TS Changes (Markups) for SQN Units 1 and 2 2.

Proposed TS Changes (Markups) for WBN Units 1 and 2 3.

Proposed TS Bases Changes (Markups) for SQN Units 1 and 2 (For Information Only) 4.

Proposed TS Bases Changes (Markup) for WBN Units 1 and 2 (For Information Only)

Enclosure CNL-24-021 E1 of 11 Description and Assessment of the Proposed Change 1.0

SUMMARY

DESCRIPTION In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, Application for amendment of license, construction permit, or early site permit, Tennessee Valley Authority (TVA) is submitting a request for an amendment to Renewed Facility Operating License Nos. DPR-77 and DPR-79 for Sequoyah Nuclear Plant (SQN), Units 1 and 2, and Facility Operating License Nos. NPF-90 and NPF-96 for Watts Bar Nuclear Plant (WBN), Units 1 and 2.

The proposed license amendment would revise SQN Units 1 and 2 and WBN Units 1 and 2 Technical Specification (TS) Limiting Condition of Operation (LCO) 3.5.2, ECCS - Operating, Note 1 to include the residual heat removal (RHR) pump flow paths as follows.

In MODE 3, both safety injection (SI) pump flow paths and one residual heat removal (RHR) pump flow path may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform pressure isolation valve testing per SR 3.4.14.1.

provides a marked-up version of the affected technical specification (TS) pages of SQN Units 1 and 2 showing the proposed changes. Attachment 2 provides a marked-up version of the affected TS pages of WBN Units 1 and 2 showing the proposed changes. provides a marked-up version of the SQN Units 1 and 2 TS Bases showing the proposed changes. Attachment 4 provides a marked-up version of the WBN Units 1 and 2 TS Bases showing the proposed changes. Changes to the existing TS Bases are provided for information only and will be implemented under the Technical Specification Bases Control Program.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation The emergency core cooling system (ECCS) at SQN and WBN consists of three separate subsystems: centrifugal charging (CC) (high head), SI (intermediate head) and RHR (low head).

Each of these subsystems consists of two independent and redundant 100 percent capacity trains. The function of the ECCS is to provide core cooling and negative reactivity to ensure that the reactor core is protected after any of the following accidents:

a) A pipe break or spurious valve lifting in the reactor coolant system (RCS) which cause a discharge larger than that which can be made up by the normal makeup system, up to and including the instantaneous circumferential rupture of the largest pipe in the RCS.

b) Rupture of a control rod drive mechanism causing a rod cluster control assembly ejection accident.

c) A pipe break or spurious valve lifting in the secondary system, up to and including the instantaneous circumferential rupture of the largest pipe in the secondary system.

d) A steam generator tube rupture (SGTR).

The analyses and acceptance criteria for the consequences of each of these accidents is described in the respective accident analyses sections of the SQN and WBN Updated Final Safety Analysis Report (UFSAR), Chapter 15. The accident analysis specified by 10 CFR 50.46

Enclosure CNL-24-021 E2 of 11 "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors," is presented in Section 15.4.1 for both SQN and WBN and shows compliance with the acceptance criteria; that is;

1. Peak cladding temperature. The calculated maximum fuel element cladding temperature shall not exceed 2200°F.

2. Maximum cladding oxidation. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

3. Maximum hydrogen generation. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

4. Coolable geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

5. Long-term cooling. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

In addition to compliance with 10 CFR 50.46, the SQN and WBN ECCS is designed to comply with General Design Criteria 35 of the 10 CFR 50, Appendix A and is discussed in Section 3.1 of the SQN and WBN UFSAR. Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities are provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure. During plant shutdown conditions, full ECCS capability may not be available and manual actuation of the required ECCS is used to mitigate the consequences of a design basis accident as described in the SQN and WBN TS Bases for LCO 3.5.3.

The proposed changes to SQN and WBN TS LCO 3.5.2 Note 1 provide equivalent exceptions to operability requirements for ECCS components of both SQN and WBN. In addition, the proposed changes maintain the compliance of SQN and WBN ECCS with the requirements of 10 CFR 50, Appendix A, and 10 CFR 50.46.

2.2 Reason for the Proposed Change The proposed TS change is being requested for both SQN and WBN in order to provide allowance within LCO 3.5.2 Note 1 regarding the ability to perform reactor coolant pressure isolation valve leakage surveillances for a period of up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months while in Mode 3. Specifically, surveillance requirement (SR) 3.4.14.1 requires verification that leakage from each RCS pressure isolation valve is within its limits when the RCS pressure is greater than or equal to 2215 pounds per square inch (psig) and less than or equal to 2255 psig. The surveillance is performed at lower RCS pressures and, therefore, a calculated extrapolation is necessary in order to determine the projected leak rate at full RCS pressures. The surveillance is currently performed in Mode 4 and testing of the RHR cold leg RCS pressure isolation valves while in Mode 3 allows for valve seating forces that are more closely representative of the RCS pressure conditions required by the TS. This aids in prevention of in-leakage into the surveillance test

Enclosure CNL-24-021 E3 of 11 boundary at lower test pressures which is included in the extrapolation to the full pressure leakage value.

The performance of SR 3.4.14.1 supports LCO 3.5.2 to ensure that two ECCS trains are operable in Modes 1, 2, and 3. Figure 1 is a simplified diagram that depicts the ECCS system of both SQN and WBN. The SI and RHR pumps take suction from the refueling water storage tank (RWST), which is not shown. The RCS pressure isolation valves being tested according to SR 3.4.14.1 are the check valves upstream of each of the four RCS cold legs on the left side of Figure 1. Fully isolating the test boundary for performing SR 3.4.14.1 for the RHR cold leg pressure isolation valves requires both SI pump flow paths and one RHR pump injection flow path to be isolated while in Mode 3. As shown in Figure 1, the SI pumps can be isolated with the closure of a single valve labeled 63-22. The isolation valves of the RHR pump flow paths are labeled 63-93 and 63-94 and separate the RHR injection flow paths into two of the RCS cold legs. Closure of the RHR pump flow path isolations are alternated when performing SR 3.4.14.1.

Figure 1 Simplified Flow Diagram of ECCS Safety Injection and RHR

Enclosure CNL-24-021 E4 of 11 The SR 3.4.14.1 leakage surveillance testing of the RCS cold leg pressure isolation valves requires both SI pump flow paths and one of the RHR pump injection flow paths into the cold legs to be isolated. The feasibility of performing this surveillance with the injection flow isolation valves in the open position has been evaluated with the following concerns identified.

1. Other systems could leak into the check valve test lines and cause false readings of check valve leakage.

2. The time required to perform the SR 3.4.14.1 surveillance would be significantly increased.

3. If check valve leakage is detected, it would be difficult to determine which of several valves is leaking.

For these reasons, a change to TS LCO 3.5.2 Note 1 is desired in order to facilitate isolation of both SI pump and one of the RHR pump injection flow paths while in Mode 3 for surveillance testing optimization.

SQN Units 1 and 2 initially requested alternate wording to LCO 3.5.2 Note 1 as part of the improved TS conversion, as described in Nuclear Regulatory Commission (NRC) Accession Number ML13330A929. The requested wording was to allow for isolation of both ECCS pump flow paths for a duration of up to two hours while in Mode 3. This was identified as a less restrictive change from the TS requirements which allowed for isolation of only the safety injection pump flow paths in Mode 3. The NRC requested additional information regarding ECCS performance in the event of a loss of cooling accident (LOCA) in Mode 3 and limitations on performing simultaneous pressure isolation valve surveillances. SQN subsequently withdrew the proposed change and implemented TS wording for LCO 3.5.2 Note 1 that is consistent with NUREG-1431 standard technical specifications for Westinghouse plants which limits the allowance only to SI pump flow paths. This license amendment request addresses the NRC requests for additional information provided during the Sequoyah Units 1 and 2 improved TS conversion.

2.3 Description of the Proposed Change The proposed license amendment request would revise the following SQN Units 1 and 2 and WBN Units 1 and 2 TS LCO 3.5.2, ECCS - Operating, Note 1 to include the RHR pump flow paths as follows:

In MODE 3, both safety injection (SI) pump flow paths and one residual heat removal (RHR) pump flow path may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform pressure isolation valve testing per SR 3.4.14.1.

provides a marked-up version of the affected TS pages of SQN Units 1 and 2 showing the proposed changes. Attachment 2 provides a marked-up version of the affected TS pages of WBN Units 1 and 2 showing the proposed changes. Attachment 3 provides a marked-up version of the SQN Units 1 and 2 TS Bases. Attachment 4 provides a marked-up version of the WBN Units 1 and 2 TS Bases. Changes to the existing TS Bases are provided for information only and will be implemented under the Technical Specification Bases Control Program. There are no changes to the SQN or WBN in-service testing program requirements as a result of this proposed amendment.

Enclosure CNL-24-021 E5 of 11

3.0 TECHNICAL EVALUATION

The amendment being requested would allow both SI pump flow paths and either of the RHR pump flow paths to be isolated for a period of up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months while in Mode 3 to perform leakage testing of RCS pressure isolation check valves per SR 3.4.14.1. A primary consideration regarding the proposed change is the ability of the ECCS system to respond to a LOCA while the RCS pressure isolation check valve leakage surveillance is being performed since the test configuration would temporarily limit the ability of ECCS to mitigate the effects of a LOCA.

A large spectrum of LOCA break sizes has been evaluated in the UFSAR accident analysis for Mode 1 conditions. No Mode 3 analyses are reported in the UFSAR, as the conditions are assumed to be bounded by the Mode 1 LOCA analyses. However, various shutdown LOCA studies have been performed in support of emergency operating procedure development and licensing applications. Given that the majority of Mode 3 is spent at lower temperatures and pressures, probabilistic fracture mechanics limits the fault sizes that could occur when the pressure isolation valve surveillance is performed during Mode 3. As such, the Mode 3 LOCA evaluations considered demonstrate that a large break LOCA with a diameter greater than six inches is not credible and for all practical purposes can be assumed not to occur. Therefore, for purposes of this discussion large break LOCAs were not necessary for evaluation.

A shutdown LOCA scoping study was performed by Westinghouse for a four-loop plant two hours after a reactor trip (ML010920123). Based on this study, it was estimated that at least 20 minutes would be available to initiate SI flow from a centrifugal charging pump to prevent significant core uncovery for breaks up to three inches in diameter. Initiation of SI flow within this time may not preclude core uncovery, but is expected to limit the fuel cladding heatup to less than the full power small break LOCA results provided in the UFSAR. For breaks larger than three inches and up to six inches in diameter, operator action to initiate SI from a centrifugal charging pump is estimated to be required within approximately 10 minutes.

Depending upon break size, additional operator action can be taken within one hour of the event initiation to start an additional centrifugal charging pump or SI pump or depressurize the RCS using the steam generators and to start an RHR pump.

More recent shutdown LOCA studies have been performed with similar conclusions.

Specifically, the Westinghouse owners group evaluated plant shutdown LOCA responses in WCAP-12476 in the early 1990s. This work also concluded that the possibility of a large break LOCA occurring during shutdown is extremely remote and that a small break LOCA would remain within 10 CFR 50.46 acceptance criteria if injection from both a high head and a low head SI pump were manually initiated within 10 minutes after the appropriate symptoms were reached. Although WCAP-12476 was withdrawn without review, the Pressurized Water Reactor Owners Group (PWROG) later performed PWROG-19021 in support of the development of PWR Emergency Response Guidelines, Revision 4. The result was the conclusion that for a Westinghouse four-loop nuclear steam supply system (NSSS), the initiation of a safety injection within 10 minutes of indication of the LOCA precluded any meaningful reactor core uncovery and, as a result, no significant core heatup occurred and the 10 CFR 50.46 criteria would be met.

There are several factors which provide assurance that operators can identify and significantly decrease the severity of a LOCA occurring during the SR 3.4.14.1 pressure isolation valve surveillance testing. Indications are available to the operator to identify that a LOCA is in progress. These include a loss of pressurizer level, RCS pressure decrease, loss of RCS subcooling, radiation alarms inside containment, containment pressure increase and sump water level increase. When a LOCA has been identified the operator would manually initiate a SI as required by plant emergency operating procedures. In addition, the emergency operating

Enclosure CNL-24-021 E6 of 11 response procedures direct the operator to verify ECCS flow. Since these valves would only be closed for up to a two hour period, the operators would be aware these valves are closed and would open them to restore all available ECCS flow. The SI and RHR pump flow paths isolation valves are readily restorable from the control room.

Because of the plant configuration during the RCS pressure isolation valve leakage surveillance, when a SI is initiated the ECCS flow would automatically be available from two centrifugal charging pumps injecting into all four cold legs and one RHR pump initially injecting into two of the cold legs. For both WBN and SQN, in Mode 3 the RHR pump suction would be aligned to the RWST. If the pressure isolation valve surveillance test were to occur in Mode 3 at an RCS pressure of 1000 psi or greater then the accumulators are operable and available to inject during a LOCA. If the pressure isolation valve surveillance were to occur at an RCS pressure of less than 1000 psi then the accumulators cannot be assumed to be available.

Regardless of availability of the accumulators, at both WBN and SQN substantially more ECCS flow would be provided in ten minutes due to the availability of all the ECCS pumps and injection paths than assumed in the various Mode 3 LOCA evaluations which do not credit accumulator injection at reduced RCS pressures.

The various Mode 3 LOCA evaluations assumed the event occurs between one and four hours after a reactor trip. The SR 3.4.14.1 RHR pressure isolation valve surveillances are most often performed during startup following a plant shutdown. The SR 3.4.14.1 RHR pressure isolation valve surveillance is only to be performed when plant conditions have stabilized. Therefore, the surveillance would be conducted when the initial fuel rod temperature and decay heat levels are less than those assumed in the various Mode 3 LOCA evaluations.

In addition to providing low head safety injection, the RHR subsystem supplements the containment spray system by providing a spray of borated water into the upper region of containment through a separate RHR spray ring header. This safety-related function of RHR limits the rise in containment pressure and temperature during a design basis accident. The RHR containment spray valves required to be operable and aligned for RHR to perform this safety-related function are not affected by performance of SR 3.4.14.1. RHR spray is not established until at least one hour after event initiation. Therefore, the ability of the RHR subsystem to support the containment spray safety function of limiting rise in containment pressure and temperature is not affected by this proposed change.

If implemented, this change will maintain the overall margin of safety for both SQN Units 1 and 2 and WBN Units 1 and 2 which is accomplished by offsetting a temporary delay in delivery of a portion of the ECCS injection flow against the assurance of maintaining the integrity of the RHR pump flow path RCS pressure isolation valves by performing the SR 3.4.14.1 surveillance testing. The probability of an intersystem LOCA initiated by the failure of these RHR pump flow path isolation valves is reduced while operating at higher plant Modes through the performance of the required surveillance.

The current TS LCO 3.5.2 Note 2 for SQN Units 1 and 2 and WBN Units 1 and 2 provides an allowance for temporary isolation of ECCS pumps in Mode 3 which is similar to that being requested. SQN Units 1 and 2 LCO 3.5.2 Note 2 allows for ECCS pumps to be made incapable of injecting to support transition into or from the applicability of LCO 3.4.12, Low Temperature Overpressure Protection (LTOP) System, for up to four hours or until the temperature of the RCS cold legs exceeds the LTOP arming temperature. Further, WBN Units 1 and 2 TS LCO 3.5.2 Note 2 allows for the SI and charging pumps to be made incapable of injection for up to four hours until the cold overpressure mitigation system (COMS) arming temperature is exceeded in the four RCS cold legs. Although the intent of LCO 3.5.2 Note 1 and Note 2 are different, the action of isolating ECCS pumps from their ability to inject into the RCS while in

Enclosure CNL-24-021 E7 of 11 Mode 3 is the same. Further, the four hour timeframe provided by LCO 3.5.2 Note 2 to realign ECCS injection while the plant is in Mode 3 conditions is bounding of the two hour timeframe being proposed by this license amendment request to allow for surveillance of RHR safety related RCS pressure isolation valves while in Mode 3.

In summary, a LOCA occurring during Mode 3 conditions at reduced RCS pressure is highly unlikely. A LOCA occurring in Mode 3 during the short period of time that the RHR pump flow path isolation valves are closed concurrently is even more unlikely. However, if a LOCA did occur, the preceding discussion demonstrates that sufficient ECCS flow would be available to maintain 10 CFR 50.46 compliance. Additionally, during a small break LOCA sufficient time exists for operator response to manually align the isolated ECCS pump flow paths. The ability of the RHR subsystem to support the containment spray safety function of limiting rise in containment pressure and temperature is not affected by this proposed change. This proposed change will maintain the margin of safety for both SQN Units 1 and 2 and WBN Units 1 and 2 by offsetting a temporary delay in delivery of RHR injection flow to two cold legs against the assurance of maintaining the integrity of the RHR pump flow path isolation valves. Finally, both SQN Units 1 and 2 and WBN Units 1 and 2 currently have a similar allowance in TS LCO 3.5.2 Note 2 to make ECCS pumps incapable of injection while in Mode 3 for a timeframe greater than that being requested.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Criteria SQN Units 1 and 2 were designed to meet the intent of the "Proposed General Design Criteria (GDC) for Nuclear Power Plant Construction Permits published in July 1967. The SQN construction permit was issued in May 1970. The UFSAR, however, addresses the Nuclear Regulatory Commission (NRC) GDC published as Appendix A to 10 CFR 50 in July 1971.

WBN Units 1 and 2 were designed to meet the intent of the "Proposed General Design Criteria for Nuclear Power Plant Construction Permits" published in July 1967. The WBN construction permit was issued in January 1973. The dual-unit UFSAR, however, addresses the NRC GDC published as Appendix A to 10 CFR 50 in July 1971.

Each criterion listed below is followed by a discussion of the design features and procedures of SQN Units 1 and 2 and WBN Units 1 and 2 that meet the intent of the criteria.

Criterion 35 - Emergency Core Cooling. A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Enclosure CNL-24-021 E8 of 11 Compliance The proposed change does not alter the ability for the emergency core cooling functions to actuate or perform its safety function. The proposed change is consistent with the SQN and WBN design and analysis which ensures sufficient and redundant heat transfer from the reactor core following a loss of reactor coolant. Therefore, the recommendations of GDC 35 continue to be met with the proposed change. Further discussion of compliance with GDC 35 is provided in SQN UFSAR Section 3.1.2 and WBN UFSAR Section 3.1.2.3.

Criterion 36 - Examination of Emergency Core Cooling. The ECCS shall be designed to permit appropriate periodic examination of important components, such as spray rings in the reactor pressure vessel, water injection nozzles, and piping, to assure the integrity and capability of the system.

Compliance The proposed change does not affect the ability to perform examination of the ECCS components. Therefore, the recommendations of GDC 36 continue to be met with the proposed change. Further discussion of compliance with GDC 36 is provided in SQN UFSAR Section 3.1.2 and WBN UFSAR Section 3.1.2.3.

Criterion 37 - Testing of Emergency Core Cooling. The ECCS shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the system as a whole and under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the Protection System, the transfer between normal and emergency power sources, and the operation of the Associated Cooling Water System.

Compliance The proposed change does not alter the ability to test the emergency core cooling functions.

Proof tests of the ECCS components are performed in the manufacturer's shop. Preoperational system hydrostatic and performance tests demonstrate structural and leaktight integrity of ECCS components and proper functioning of the system. Thereafter, periodic tests demonstrate that the system components are functioning properly. The proposed change is consistent with the SQN and WBN design and supports the performance of adequate leaktight integrity of the RHR subsystem. Therefore, the recommendations of GDC 37 continue to be met with the proposed change. Further discussion of compliance with GDC 37 is provided in SQN UFSAR Section 3.1.2 and WBN UFSAR Section 3.1.2.3.

Therefore, with the implementation of the proposed change, SQN Units 1 and 2 and WBN Units 1 and 2 continue to meet the identified applicable GDC, regulations, and requirements.

4.2 Precedent Byron Units 1 and 2 A license amendment request was submitted initially in November of 1985 for the Byron Station Units 1 and 2 which proposed a similar allowance for the isolation of RHR pump flow paths in

Enclosure CNL-24-021 E9 of 11 order to perform surveillance leakage testing of the pressure isolation valves (ML20137T478).

The request was modified and additional information was provided to the NRC in ML20206J786 and ML20214W184. The license amendment request for Byron Station Units 1 and 2 was subsequently approved by the NRC in ML020850356. The current Byron Station Units 1 and 2 TS retains the allowance to isolate RHR flow paths to perform pressure isolation valve testing while in Mode 3 (ML052910365 and ML052910368). The Byron precedent is the primary basis for the addition of LCO 3.5.2 Note 1 to the NUREG-1431 standard technical specifications and references the same shutdown LOCA scoping study performed by Westinghouse for a four-loop plant two hours after a reactor trip (ML010920123).

McGuire Units 1 and 2 The McGuire Units 1 and 2 current TS (ML052870418 and ML052870419) include wording in LCO 3.5.2 Note 1 that is encompassing of that being requested for SQN Units 1 and 2 and WBN Units 1 and 2. The change for McGuire Units 1 and 2 was requested as part of conversion to the improved TS format of NUREG-1431. The change was identified by McGuire in ML20248L685 as being more restrictive than the TS requirements in place at the time because of the limitation of the pressure isolation valve realignments to Mode 3 with a two hour limited time duration. The prior McGuire Units 1 and 2 TS placed no time limit on the pressure isolation valve testing.

Vogtle Units 1 and 2 The current TS for Vogtle Units 1 and 2 (ML052840233 and ML052840236) include wording in LCO 3.5.2 Note 1 that allow for either of the RHR pump to cold leg injection flow paths to be isolated for the performance of RCS pressure isolation valve surveillance. Vogtle identified this as a conservative variation from the NUREG-1431 TS format since the licensing basis at the time did not impose the two hour time limit on the RHR flow path isolations (ML20095K940). At least part of the basis for this allowance was identified by Vogtle to be contained in an NRC staff evaluation from November of 1987 (ML20236K912).

4.3 No Significant Hazards Consideration Determination Analysis In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, Application for amendment of license, construction permit, or early site permit, Tennessee Valley Authority (TVA) is submitting a request for an amendment to Renewed Facility Operating License Nos. DPR-77 and DPR-79 for Sequoyah Nuclear Plant (SQN), Units 1 and 2, and Facility Operating License Nos. NPF-90 and NPF-96 for Watts Bar Nuclear Plant (WBN), Units 1 and 2.

The proposed license amendment would revise the following to SQN Units 1 and 2 and WBN Units 1 and 2 Technical Specification (TS) Limiting Condition of Operation (LCO) 3.5.2, ECCS - Operating, Note 1 to include residual heat removal (RHR) pump flow paths as follows.

In MODE 3, both safety injection (SI) pump flow paths and one residual heat removal (RHR) pump flow path may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform pressure isolation valve testing per SR 3.4.14.1.

TVA has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below.

Enclosure CNL-24-021 E10 of 11

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

Response: No The conclusion from the review of the Updated Final Safety Analysis Report analysis is that there is no significant increase in the probability or consequence of any previously evaluated accident. The RHR pump flow path isolation valves would only be closed for the short period of time required to perform the pressure isolation valve surveillance. The probability of a loss of cooling accident (LOCA) occurring during this short duration is highly unlikely and the dose consequences of such an event remain unchanged. In the event of a LOCA, these valves are readily operable from the control room and the operator would open them as required by the Emergency Operating Procedures. The technical evaluation provided demonstrates that sufficient emergency core cooling systems flow would be available to maintain 10 CFR 50.46 compliance.

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

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

Response: No.

The postulated accident that is potentially affected by a reduction in delivered RHR flow is a LOCA. This change only affects the temporary closure of RHR pump isolation valves to allow for reactor coolant system leakage testing for a short period of time. If a LOCA did occur, the technical evaluation provided demonstrates that sufficient emergency core cooling systems flow would be available to maintain 10 CFR 50.46 compliance.

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

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

Response: No.

The proposed amendment does not involve a significant reduction in a margin of safety. Rather, this change will maintain the margin of safety for SQN Units 1 and 2 and WBN Units 1 and 2. This is accomplished by offsetting a temporary delay in the delivery of a portion of ECCS flow during the improbable occurrence of a LOCA occurring during the short duration of a surveillance which provides assurance of pressure isolation valve integrity. The change increases the efficiency of the testing and enhances the test configuration.

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

Based on the above, TVA concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

Enclosure CNL-24-021 E11 of 11 4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) 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.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Enclosure CNL-24-021 Proposed Technical Specification Changes (Mark-Up) for SQN Units 1 and 2

ECCS - Operating 3.5.2 SEQUOYAH - UNIT 1 3.5.2-1 Amendment 358,334, 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

NOTES-------------------------------------------

1.

In MODE 3, both safety injection (SI) pump flow paths and one residual heat removal (RHR) pump flow path may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform pressure isolation valve testing per SR 3.4.14.1.

2.

In MODE 3, ECCS pumps may be made incapable of injecting to support transition into or from the Applicability of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or until the temperature of all RCS cold legs exceeds Low Temperature Overpressure Protection (LTOP) arming temperature specified in the PTLR plus 25°F, whichever comes first.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains inoperable.

A.1 Restore train(s) to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OR In accordance with the Risk Informed Completion Time Program B. Required Action and associated Completion Time not met.

B.1 Be in MODE 3.

AND B.2 Be in MODE 4.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours C. Less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.

C.1 Enter LCO 3.0.3.

Immediately

ECCS - Operating 3.5.2 SEQUOYAH - UNIT 2 3.5.2-1 Amendment 352, 327, 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

NOTES-------------------------------------------

1.

In MODE 3, both safety injection (SI) pump flow paths and one residual heat removal (RHR) pump flow path may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform pressure isolation valve testing per SR 3.4.14.1.

2.

In MODE 3, ECCS pumps may be made incapable of injecting to support transition into or from the Applicability of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or until the temperature of all RCS cold legs exceeds Low Temperature Overpressure Protection (LTOP) arming temperature specified in the PTLR plus 25°F, whichever comes first.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains inoperable.

A.1 Restore train(s) to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OR In accordance with the Risk Informed Completion Time Program B. Required Action and associated Completion Time not met.

B.1 Be in MODE 3.

AND B.2 Be in MODE 4.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours C. Less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.

C.1 Enter LCO 3.0.3.

Immediately

Enclosure CNL-24-021 Proposed Technical Specification Changes (Mark-Up) for WBN Units 1 and 2

ECCS - Operating 3.5.2 Watts Bar-Unit 1 3.5-4 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

NOTES------------------------------------------

In MODE 3, both safety injection (SI) pump flow pathsDQGRQHresidual heat removal (5+5) SXPSIORZSDWKmay be isolatedby closing the isolation valves for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months toperform pressureisolation valve testing per SR 3.4.14.1.

In MODE 3, the safety injection pumps and charging pumps may be

made incapable of injecting to support transition into or from the

Applicability of the LCO 3.4.12, Cold Overpressure Mitigation System

(COMS) for up to four hours or until the temperature of all the RCS cold

legs exceeds 375qF, whichever occurs first.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more trains inoperable.

AND At least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.

A.1 Restore train(s) to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B.

Required Action and associated Completion Time not met.

B.1 Be in MODE 3.

AND B.2 Be in MODE 4.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours Amendment ,

ECCS - Operating 3.5.2 Watts Bar - Unit 2 3.5-3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 APPLICABILITY:

Two ECCS trains shall be OPERABLE.

NOTES------------------------------------------

In MODE 3, both safety injection (SI) pump flow pathsDQGRQH

residual heat removal (RHR) SXPSIORZSDWKmay be isolatedby closing the isolation valves for up to2 hours toperform pressureisolation valve testing per SR 3.4.14.1.

In MODE 3, the safety injection pumps and charging pumps may be

made incapable of injecting to support transition into or from the

Applicability of the LCO 3.4.12, Cold Overpressure Mitigation System (COMS) for up to four hours or until the temperature of all the RCS cold

legs exceeds the COMS arming temperature specified in the PTLR,

whichever occurs first.

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains inoperable.

AND At least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.

A.1 Restore train(s) to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B. Required Action and associated Completion Time not met.

B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND B.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Amendment

Enclosure CNL-24-021 Proposed Technical Specification Bases Changes (Mark-Up) for SQN Units 1 and 2 (For Information Only)

ECCS - Operating B 3.5.2 SEQUOYAH - UNIT 1 B 3.5.2-5 Revision 45

BASES LCO (continued)

In MODES 1, 2, and 3, an ECCS train consists of a centrifugal charging subsystem, an SI subsystem, and an RHR subsystem. Each ECCS train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST upon an SI signal and automatically transferring RHR suction to the containment sump.

During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the RWST to the RCS via the ECCS pumps and their respective supply headers to each of the four cold leg injection nozzles. In the long term, this flow path may be switched to take its supply from the containment sump and to supply its flow to the RCS hot and cold legs.

The flow path for each ECCS train must maintain its designed independence to ensure that no single failure can disable both ECCS trains.

As indicated in Note 1, the SI pump flow paths and one RHR pump flow SDWKmay be isolated for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in MODE 3, under controlled conditions, to perform pressure isolation valve testing per SR 3.4.14.1. The flow path is readily restorable from the control room.

As indicated in Note 2, operation in MODE 3 with ECCS trains made incapable of injecting in order to facilitate entry into or exit from the Applicability of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," is necessary for plants with an LTOP arming temperature at or near the MODE 3 boundary temperature of 350°F.

LCO 3.4.12 requires that certain pumps be rendered incapable of injecting at and below the LTOP arming temperature. When this temperature is at or near the MODE 3 boundary temperature, time is needed to make pumps incapable of injecting prior to entering the LTOP Applicability, and provide time to restore the inoperable pumps to OPERABLE status on exiting the LTOP Applicability.

APPLICABILITY In MODES 1, 2, and 3, the ECCS OPERABILITY requirements for the limiting Design Basis Accident, a large break LOCA, are based on full power operation. Although reduced power would not require the same level of performance, the accident analysis does not provide for reduced cooling requirements in the lower MODES. The centrifugal charging pump performance is based on a small break LOCA, which establishes the pump performance curve and has less dependence on power. The SI pump performance requirements are based on a small break LOCA.

MODE 2 and MODE 3 requirements are bounded by the MODE 1 analysis.

ECCS - Operating B 3.5.2 SEQUOYAH - UNIT 2 B 3.5.2-5 Revision 45