ML24218A144
| ML24218A144 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 08/27/2024 |
| From: | Kimberly Green Plant Licensing Branch II |
| To: | Jim Barstow Tennessee Valley Authority |
| Green K | |
| References | |
| EPID L-2023-LLA-0152 | |
| Download: ML24218A144 (21) | |
Text
August 27, 2024 James Barstow Vice President, Nuclear Regulatory Affairs and Support Services Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801
SUBJECT:
WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 169 AND 75 REGARDING TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT 3.9.5.1 TO REDUCE THE RESIDUAL HEAT REMOVAL FLOW RATE (EPID L-2023-LLA-0152)
Dear James Barstow:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 169 to Facility Operating License No. NPF-90 and Amendment No. 75 to Facility Operating License No. NPF-96 for the Watts Bar Nuclear Plant (Watts Bar), Units 1 and 2, respectively. These amendments are in response to your application dated October 30, 2023, as supplemented by letters dated January 10, 2024, and June 27, 2024.
The amendments revise the Watts Bar, Units 1 and 2, Technical Specification Surveillance Requirement 3.9.5.1 to reduce the minimum required flow rate for circulating reactor coolant with one residual heat removal loop in operation from 2,500 gallons per minute (gpm) to 2,000 gpm.
A copy of our related safety evaluation is also enclosed. A notice of issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA Perry Buckberg Acting for/
Kimberly J. Green, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-390 and 50-391
Enclosures:
- 1. Amendment No. 169 to NPF-90
- 2. Amendment No. 75 to NPF-96
- 3. Safety Evaluation cc: Listserv
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-390 WATTS BAR NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 169 License No. NPF-90
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (TVA), the licensee, dated October 30, 2023, as supplemented by letters dated January 10, 2024, and June 27, 2024, 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 Facility Operating License No. NPF-90 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 169 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 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 David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Operating License and Technical Specifications Date of Issuance: August 27, 2024 BLAKE PURNELL Digitally signed by BLAKE PURNELL Date: 2024.08.27 14:57:11 -04'00'
ATTACHMENT TO AMENDMENT NO. 169 WATTS BAR NUCLEAR PLANT, UNIT 1 FACILITY OPERATING LICENSE NO. NPF-90 DOCKET NO. 50-390 Replace page 3 of Facility Operating License No. NPF-90 with the attached revised page 3. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Replace the following page of the Appendix A, Technical Specifications, with the attached page.
The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Page Insert Page 3.9-9 3.9-9
Amendment No. 169 Facility License No. NPF-90 (4)
TVA, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required, any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis, instrument calibration, or other activity associated with radioactive apparatus or components; and (5)
TVA, pursuant to the Act and 10 CFR Parts 30, 40 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 license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I 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 TVA is authorized to operate the facility at reactor core power levels not in excess of 3459 megawatts thermal.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 169 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Safety Parameter Display System (SPDS) (Section 18.2 of SER Supplements 5 and 15)
Prior to startup following the first refueling outage, TVA shall accomplish the necessary activities, provide acceptable responses, and implement all proposed corrective actions related to having the Watts Bar Unit 1 SPDS operational.
(4)
Vehicle Bomb Control Program (Section 13.6.9 of SSER 20)
During the period of the exemption granted in paragraph 2.D.(3) of this license, in implementing the power ascension phase of the approved initial test program, TVA shall not exceed 50% power until the requirements of 10 CFR 73.55(c)(7) and (8) are fully implemented. TVA shall submit a letter under oath or affirmation when the requirements of 73.55(c)(7) and (8) have been fully implemented.
RHR and Coolant Circulation - High Water Level 3.9.5 Watts Bar-Unit 1 3.9-9 Amendment 132, 169 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
(continued)
A.4 Close all containment penetrations providing direct access from containment atmosphere to outside atmosphere.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1 Verify one RHR loop is in operation and circulating reactor coolant at a flow rate of 2000 gpm.
In accordance with the Surveillance Frequency Control Program
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-391 WATTS BAR NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 75 License No. NPF-96
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (TVA), the licensee, dated October 30, 2023, as supplemented by letters dated January 10, 2024, and June 27, 2024, 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 Facility Operating License No. NPF-96 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 75 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 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 David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Operating License and Technical Specifications Date of Issuance: August 27, 2024 BLAKE PURNELL Digitally signed by BLAKE PURNELL Date: 2024.08.27 14:58:03 -04'00'
ATTACHMENT TO AMENDMENT NO. 75 WATTS BAR NUCLEAR PLANT, UNIT 2 FACILITY OPERATING LICENSE NO. NPF-96 DOCKET NO. 50-391 Replace page 3 of Facility Operating License No. NPF-96 with the attached revised page 3. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Replace the following page of the Appendix A, Technical Specifications, with the attached page.
The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Page Insert Page 3.9-7 3.9-7 Unit 2 Facility Operating License No. NPF-96 Amendment No. 75 C.
The license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I 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 TVA is authorized to operate the facility at reactor core power levels not in excess of 3459 megawatts thermal.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 75 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
TVA shall implement permanent modifications to prevent overtopping of the embankments of the Fort Loudon Dam due to the Probable Maximum Flood by June 30, 2018.
(4)
FULL SPECTRUM LOCA Methodology shall be implemented when the WBN Unit 2 steam generators are replaced with steam generators equivalent to the existing steam generators at WBN Unit 1.
(5)
By December 31, 2019, the licensee shall report to the NRC that the actions to resolve the issues identified in Bulletin 2012-01, Design Vulnerability in Electrical Power System, have been implemented.
(6)
The licensee shall maintain in effect the provisions of the physical security plan, security personnel training and qualification plan, and safeguards contingency plan, and all amendments made pursuant to the authority of 10 CFR 50.90 and 50.54(p).
(7)
TVA shall fully implement and maintain in effect all provisions of the Commission approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).
The TVA approved CSP was discussed in NUREG-0847, Supplement 28, as amended by changes approved in License Amendment No. 7.
(8)
TVA shall implement and maintain in effect all provisions of the approved fire protection program as described in the Fire Protection Report for the facility, as described in NUREG-0847, Supplement 29, subject to the following provision:
RHR and Coolant Circulation - High Water Level 3.9.5 Watts Bar - Unit 2 3.9-7 Amendment 36, 75 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued)
A.4 Close all containment penetrations providing direct access from containment atmosphere to outside atmosphere.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1 Verify one RHR loop is in operation and circulating reactor coolant at a flow rate of 2000 gpm.
In accordance with the Surveillance Frequency Control Program
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 169 TO FACILITY OPERATING LICENSE NO. NPF-90 AND AMENDMENT NO. 75 TO FACILITY OPERATING LICENSE NO. NPF-96 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-390 AND 50-391
1.0 INTRODUCTION
By letter dated October 30, 2023 (Agencywide Documents Access and Management System Accession No. ML23303A095), as supplemented by letters dated January 10, 2024 (ML24010A064), and June 27, 2024 (ML24179A207), the Tennessee Valley Authority (TVA or the licensee) submitted a license amendment request (LAR) for Watts Bar Nuclear Plant (Watts Bar), Units 1 and 2. The requested changes would revise the Watts Bar, Units 1 and 2, Technical Specification (TS) Surveillance Requirement (SR) 3.9.5.1 to reduce the minimum required flow rate for circulating reactor coolant with one residual heat removal (RHR) loop in operation from 2,500 gallons per minute (gpm) to 2,000 gpm.
The supplemental letter dated January 10, 2024, corrected inconsistent proprietary markings in the enclosures to the LAR, but did not revise the content of the enclosures. The supplemental letter dated June 27, 2024, provided additional information that clarified the application, but did not expand the scope of the application as originally noticed, nor did it change the U.S. Nuclear Regulatory Commission (the NRC or Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on January 2, 2024 (89 FR 110).
2.0 REGULATORY EVALUATION
2.1
System Description
In Mode 6 (i.e., Refueling), the RHR system maintains the reactor coolant system (RCS) temperature by transferring heat to the component cooling system while providing mixing for borated reactor coolant, also preventing boron stratification. The RHR system also transfers water from the refueling cavity and the refueling water storage tank to support Mode 6 activities.
Additionally, the RHR system serves as part of the emergency core cooling system during the injection and recirculation phases of an accident.
Each unit has an RHR system which consists of two heat exchangers, two pumps, and associated piping, valves, and instrumentation. The RHR system takes suction from the hot leg of one reactor coolant loop, through the RHR pumps, into the tube side of the RHR heat exchangers and discharges into the cold leg of each loop. During long-term recirculation mode following a loss-of-coolant accident, the RHR system can take suction from the containment sump to cool it through the RHR heat exchangers and return it to the reactor vessel through the charging and safety injection pumps. In Mode 6, only one RHR system loop is required for decay heat removal with the reactor water level 23 feet above the reactor vessel flange.
Watts Bar, Units 1 and 2, TS 3.9.5, Residual Heat Removal (RHR) and Coolant Circulation -
High Water Level, provides the specifications for operation of RHR system during Mode 6 when the water level 23 feet above the reactor vessel flange. The limiting condition for operation (LCO) requires that one RHR loop shall be operable and in operation.
2.2 Proposed Change to Technical Specifications As described in the LAR and shown below, the licensee proposed a revision to TS SR 3.9.5.1 to reduce the minimum required flow rate for RHR (new text shown in bold, and deleted text shown in strikeout).
Verify one RHR loop is in operation and circulating reactor coolant at a flow rate of 2500 2000 gpm.
2.3 Applicable Regulatory Requirements and Guidance Under Title 10 of the Code of Federal Regulations (10 CFR) 50.90, Application for amendment of license, construction permit, or early site permit, whenever a holder of a license wishes to amend the license, including TSs in the license, an application for amendment must be filed, fully describing the changes desired. Under 10 CFR 50.92(a), determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Both the common standards for licenses in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public, and that the applicant will comply with the Commissions regulations.
Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, establishes the minimum requirements for the principal design criteria for water-cooled nuclear power plants. The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety.
According to section 3.1.1, Introduction, of the Watts Bar Dual-Unit Updated Final Safety Analysis Report (UFSAR), the plant was designed to meet the intent of the Proposed General Design Criteria for Nuclear Power Plant Construction Permits, published in July 1967. The Watts Bar construction permits were issued in January 1973. The Watts Bar plant, in general, meets the intent of the NRC GDC published as Appendix A to 10 CFR Part 50 in July 1971, as discussed in UFSAR section 3.1.2, WBNP [Watts Bar Nuclear Plant] Conformance with GDCs, (ML23346A225). The NRC staff determined that the following GDC are relevant to the review:
GDC 14, Reactor coolant pressure boundary, states that the design, fabrication, erection, and testing of the reactor coolant pressure boundary shall be so as to have an extremely low probability of abnormal leakage, or rapidly propagating failure, and of gross rupture.
GDC 15, Reactor coolant system design, states that the reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.
GDC 34, Residual heat removal, states that a system to remove residual heat shall be provided. Furthermore, that the safety function of the system shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.
The Commissions regulatory requirements related to the content of TSs are set forth in 10 CFR 50.36, Technical Specifications, which require, in pertinent part, that the TSs include: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) SRs; (4) design features; and (5) administrative controls. In accordance with 10 CFR 50.36(c)(3), SRs 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 NRC staffs guidance for the review of TSs is in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]
Edition (SRP), Chapter 16.0, Technical Specifications, Revision 3, dated March 2010 (ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared Standard Technical Specifications for each of the LWR nuclear designs.
Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with NUREG-1431,1 as modified by NRC-approved travelers.
3.0 TECHNICAL EVALUATION
The LAR included Westinghouse Electric Company LLC (Westinghouse) Letter Report, LTR-SEE-23-4-P, Revision 1, Technical Evaluation in Support of Watts Bar Units 1 & 2 Residual Heat Removal System (RHRS) Flow Rate Reduction During Mode 6 Operation at Refueling Water Level 23 Feet, as enclosure 2 (hereinafter referred to as the Letter Report).
A non-proprietary version was included as enclosure 3. The Letter Report stated that an evaluation was performed to determine the acceptability of reducing the required minimum RHR flow rate from 2,500 gpm to 2,000 gpm while in Mode 6 with the reactor vessel water level greater than or equal to 23 feet above the top of the reactor vessel flange. The Letter Report indicated that the following areas were evaluated to determine if the lower flow rate is acceptable:
- 2. The reactor coolant is mixed such that significant thermal stratification does not occur.
1 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Westinghouse Plants, NUREG-1431, Volume 1, Specifications, and Volume 2, Bases, Revision 5, September 2021 (ML21259A155 and ML21259A159, respectively).
- 3. The reactor coolant is mixed such that significant boron stratification does not occur.
- 4. The RHR bypass flow control valve (HCV-618) potential for cavitation is reduced.
- 5. The RHR System check valves are maintained in a full-open position to preclude disc chatter.
- 6. Adequate RHR pump motor thrust bearing life margin is maintained.
The Letter Report also indicated that check valve 8948 (Westinghouse valve designation) was added to the Watts Bar Inservice Inspection (ISI) Program. The corresponding TVA valve identification numbers are 1/2-CKV-63-560, 561, 562, and 563, and are also included in the Watts Bar Inservice Testing (IST) Program. The licensee stated that these valves have not experienced any RHR pump cavitation or valve chattering at the lower flow rate and water level in SR 3.9.6.1.
Based on the Letter Report and the proprietary documents referenced within said report, the licensee asserted that the applicable RHR pumps and valves at Watts Bar, Units 1 and 2, will operate satisfactorily with the reduced flow rate. The licensee determined that the reduced flow rate will have no effect on the capability of the components to operate through, and after, a seismic event. The licensee indicated that it addressed the recommendations in the Letter Report for monitoring the performance of components with the proposed minimum RHR flow rate.
The NRC staff reviewed the licensees LAR, including the proprietary Letter Report enclosed with the LAR. The staff also conducted an audit of the proprietary documents referenced in the Letter Report. The staff issued a summary report of the audit on July 25, 2024 (ML24204A265),
indicating that additional information would be needed on the docket to support the NRC staffs review of the LAR.
The NRC staff reviewed the LAR and supplements using the regulatory requirements described in section 2.3 of this safety evaluation.
3.1 Evaluation of Decay Heat Removal The Letter Report stated that the primary function of the RHR system is to remove decay heat during the second phase of a plant cooldown. As the required RHR flow rate is reduced, the capacity of the RHR system to remove decay heat is also reduced, which could lead to a longer shutdown time. This heat load decreases as the time after plant shutdown increases, which reduces the decay heat removal requirements for RHR flow. As a result, the Letter Report stated the change to the RHR system minimum flow rate to 2,000 gpm at a water level 23 feet is not a concern for decay heat removal.
According to section 5.5.7.1, Design Bases, of the Watts Bar Dual-Unit UFSAR (ML23346A225), the design heat load of the RHR system is based on the decay heat fraction that exists at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> following reactor shutdown from an extended run at full power.
Additionally, the RHR system cooldown rate is limited because, as stated in UFSAR section 5.5.7.2.2, System Operation, the initiation of the RHR system includes a warm-up period, during which time the reactor coolant flow through the heat exchangers is limited to minimize thermal shock on the heat exchangers. Once the RCS pressure is reduced and the temperature is 140 degrees Fahrenheit (°F), the reactor head may be removed for refueling or maintenance.
Because the proposed TS change applies in Mode 6 (i.e., Refueling), which occurs more than 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> following reactor shutdown, the NRC staff finds that this proposed change will not adversely affect decay heat removal.
3.2 Evaluation of Thermal Stratification The Letter Report assessed the proposed RHR system flow rate change to determine its effect on thermal stratification, or the temperature gradient across the core, because thermal stratification could negatively affect the potential for departure from nucleate boiling (DNB).
The Letter Report asserted that the potential for thermal stratification is minimized due to the design of the RHR system, where reactor coolant is mixed from two cold leg nozzles before passing through the downcomer to enter the lower plenum. However, in its submittal, the Letter Report stated:
Thus, it is possible for the local coolant temperature to exceed 200°F and approach the point of nucleate boiling. However, for the worst-case scenario evaluated, it was concluded that DNB would not be a concern at the Watts Bar Units 1 and 2 at a reduced RHR flowrate during MODE 6 operation.
The NRC staff requested additional information regarding the worst-case scenario evaluated to determine if DNB is a credible concern at the proposed reduced RHR system flow rate. In the supplemental letter dated June 27, 2024, the licensee stated that higher subcooling due to the 23-foot head above the reactor vessel, plus the 2,000 gpm flow rate (which is greater than the cooling flow analyzed in the worst-case scenario), justifies that the previous DNB analysis is conservative. The NRC staff reviewed the licensees response and determined that the proposed RHR flow of 2,000 gpm is greater than the worst-case scenario evaluated such that the existing DNB analysis with a 2,000 gpm RHR system flow rate remains valid.
Because the proposed change to the flow rate remains bounded by the existing DNB analysis, and with the increased subcooling due to the 23 feet of water present above the top of the reactor vessel flange, the NRC staff finds that the proposed change will not adversely affect the current DNB analysis due to thermal stratification.
3.3 Evaluation of Boron Mixing and Stratification One of the design functions of the RHR system is to prevent boron stratification and to keep the RCS mixed in order to ensure a uniform boron concentration. The Letter Report stated that stratification is most likely to occur following a controlled boration or dilution of the RCS when first initiated, and that the boron concentration has already been stabilized at the required shutdown margin prior to reducing the RHR flow rate. Additionally, the Letter Report stated that the refueling boron concentration is approximately 2,000 parts per million which is under 1-percent concentration. At this concentration, the saturation temperature is less than 32 °F, therefore, boron precipitation would not occur.
The NRC staff reviewed the TS bases for LCO 3.9.5 related to the control of boron levels, specifically as it relates to the TS note which allows the required operating RHR loop to be removed from service for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8-hour period provided no operations are permitted that would cause reduction of the RCS boron concentration, and verified there are no specific limits related to boron concentration specified in the TS. Licensee operation is controlled by procedures which prohibit activities that could change the boron concentration during this 1-hour period so the concentration will remain stable during this time of cooling by natural circulation.
Under the operating conditions of Mode 6, with the TS required controls on dilution events during natural circulation, the NRC staff finds that this proposed change will not adversely affect boron mixing and stratification.
3.4 Evaluation of Impact of Reduced Flow To justify the reduction of RHR flow rate, the licensee evaluated and determined, in part, that the potential for cavitation of the RHR bypass flow control valve (HCV-618) is reduced, the RHR System check valves are maintained in a full-open position to preclude disc chatter, and adequate RHR pump motor thrust bearing life margin is maintained. The licensee stated that actions were taken to accommodate low flow operations for Watts Bar, Units 1 and 2, including (a) the 10-inch check valve (8948) was monitored during the first time that RHR flow rate was reduced below 2500 gpm; and (b) check valve 8948 was added to the Watts Bar ISI Program.
The Letter Report indicated that the RHR flow rate is reduced by fully closing the RHR bypass flow control valve (HCV-618), and then slowly closing the associated hand control valve (HCV-606 or 607). The Letter Report also stated that cavitation of the reactor coolant could result when the pressure drop across the control valve (HCV-618) increases as flow is reduced.
The Letter Report noted that severe cavitation could cause excessive wear and vibration in the piping downstream of the control valve. The NRC staff requested the licensee to confirm that it had addressed these precautions. In its June 27, 2024, response, the licensee stated that the Watts Bar procedures only allow the RHR pump to be operated at greater than 2,000 gpm or secured. The procedures do not follow the recommendation in the Letter Report with respect to slowly closing valves FCV-74-28 (Westinghouse ID HCV-606) and FCV-74-16 (Westinghouse ID HCV-607); instead, the Watts Bar procedures require slowly throttling open valves FCV-74-28 and FCV-74-16 when placing RHR trains in service. When securing a train of RHR during Mode 6, valves FCV-74-28 and FCV-74-16 are closed, and not throttled, in order to avoid cavitation. The licensee also confirmed that the Watts Bar procedures address the potential for changes in check valve performance, such as chatter, and that the potential changes in performance would be detected by plant operators during routine walkdowns.
The Letter Report stated that during the first operation with reduced RHR flow rates, the 10-inch check valve 8948 was locally monitored for chatter noise. Further, the Letter Report stated that if the RHR flow rate through the valves is insufficient to maintain them in a full-open position, there is a potential for certain problems to occur. The NRC requested the licensee to confirm that potential changes in the performance of the smaller check valves (such as chatter) would be addressed during the various modes of RHR system operation. In its response dated June 27, 2024, the licensee confirmed that potential changes in the performance of check valves (such as chatter) during reduced RHR flow rates would be addressed. The licensee noted that TVA operating procedures do not allow operation below 2,000 gpm. Additionally, the licensee stated that any potential changes in the performance of check valves (such as chatter) would be detected by plant operators as part of their routine walkdowns.
Based on its review, the NRC staff finds that the licensee has provided an acceptable evaluation of the performance of the applicable pumps and valves for the proposed reduction in RHR flow rate. The staff finds that the additional actions specified by the licensee will provide adequate monitoring of the performance of those pumps and valves when the RHR flow is reduced.
Therefore, the NRC staff finds that the licensee has provided reasonable assurance of the operational readiness for the applicable pumps and valves to support the proposed reduction in RHR system minimum flow rate from 2,500 gpm to 2,000 gpm.
3.5 Technical Conclusion In summary, the NRC staff finds the proposed change to the required RHR system minimum flow rate from 2,500 gpm to 2,000 gpm acceptable because (1) the proposed change will not adversely impact the decay heat removal capability of the RHR system because the SR applies during Mode 6 in which the RHR system heat load is bounded by its design heat load; (2) thermal stratification will not be adversely impacted since the existing DNB analysis remains valid, and while this TS is applicable, there is 23 feet of water above the top of the reactor vessel flange which also increases the subcooling; (3) existing controls on boron concentration are independent of RHR system flow rate and will not be impacted by the proposed change; and (4) the reduced flow rate will not adversely impact the applicable pumps and valves.
Therefore, the NRC staff concludes that the requirements of 10 CFR 50.36(c)(3) will continue to be met because the TS SR, as amended by the proposed change, still assures that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that TS LCO 3.9.5 will be met. Specifically, the revised SR will assure that one RHR loop is in operation at the minimum flow rate of 2,000 gpm.
The staff also concludes that the RHR system operation during Mode 6 at the reduced flow will continue to meet the following GDC for the stated reasons. Specifically:
GDC 14 continues to be satisfied because the proposed change will not negatively impact the reactor coolant pressure boundary.
GDC 15 continues to be satisfied because the proposed change will not adversely affect the design of the RCS and its margin during operation.
GDC 34 continues to be satisfied because the proposed change will result in adequate decay and sensible heat removal so that fuel and pressure boundary design limits are not exceeded.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Tennessee State official was notified of the proposed issuance of the amendment on July 22, 2024. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a surveillance requirement. The NRC staff has determined that the amendments involve 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 previously issued a proposed finding that the amendment involves no significant hazards consideration published in the Federal Register on January 2, 2024 (89 FR 110), and there has been no public comment on such finding. Accordingly, the amendments meet 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 amendments.
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.
Principal Contributors: T. Scarbrough, NRR J. Ambrosini, NRR A. Sallman, NRR Date: August 27, 2024
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ALeatherman DWrona (BPurnell for)
KGreen (PBuckberg for)
DATE 08/15/2024 08/21/2024 08/27/2024 08/27/2024