ML25252A202
| ML25252A202 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  (DPR-079) |
| Issue date: | 09/22/2025 |
| From: | Kimberly Green Plant Licensing Branch II |
| To: | Erb D Tennessee Valley Authority |
| Green, K | |
| References | |
| EPID L-2025-LLA-0032 | |
| Download: ML25252A202 (1) | |
Text
September 22, 2025 Mr. Delson C. Erb Vice President, OPS Support Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801
SUBJECT:
SEQUOYAH NUCLEAR PLANT, UNIT 2 - ISSUANCE OF AMENDMENT NO. 367 REGARDING REVISION OF TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT 3.0.2 TO EXTEND ICE MASS SURVEILLANCE REQUIREMENTS ON ONE-TIME BASIS (EPID L-2025-LLA-0032)
Dear Mr. Erb:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 367 to Renewed Facility Operating License No. DPR-79, for the Sequoyah Nuclear Plant (SQN), Unit 2. The amendment is in response to your application dated February 10, 2025.
The amendment revises the SQN, Unit 2, Technical Specification Surveillance Requirement (SR) 3.0.2 to permit the extension of SRs 3.6.12.2 and 3.6.12.3 regarding ice mass surveillance on a one-time basis.
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Kimberly Green, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-328
Enclosures:
- 1. Amendment No. 367 to DPR-79
- 2. Safety Evaluation cc: Listserv
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 367 Renewed License No. DPR-79
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated February 10, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-79 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 367 are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.
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 Renewed Facility Operating License and Technical Specifications Date of Issuance: September 22, 2025 DAVID WRONA Digitally signed by DAVID WRONA Date: 2025.09.22 13:10:22 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 367 SEQUOYAH NUCLEAR PLANT, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 Replace page 3 of the Renewed Facility Operating License with the attached page 3. The revised page contains a marginal line indicating the area of change.
Replace the following pages of the Appendix A, Technical Specifications, with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 3.0-4 3.0-4 3.0-5 3.0-5 3.0-6 (3)
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 as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
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 or instrument calibration or associated with radioactive apparatus or components; and (5)
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 Sequoyah and Watts Bar Unit 1 Nuclear Plants.
C.
This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's 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 The Tennessee Valley Authority is authorized to operate the facility at reactor core power levels not in excess of 3455 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 367 are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
Initial Test Program The Tennessee Valley Authority shall conduct the post-fuel-loading initial test program (set forth in Section 14 of Tennessee Valley Authority's Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as:
- a.
Elimination of any test identified in Section 14 of TVA's Final Safety Analysis Report as amended as being essential; Amendment No. 367 Renewed License No. DPR-79
SR Applicability 3.0 SEQUOYAH - UNIT 2 3.0-4 Amendment 327, 355, 367 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR.
Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.
SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. In addition, for each of the SRs listed in Table SR 3.0.2-1 the specified Frequency is met if the Surveillance is performed on or before the Frequency extension limit.
This extension of the test intervals for these SRs is permitted on a one-time basis.
For Frequencies specified as "once," the above interval extension does not apply.
If a Completion Time requires periodic performance on a "once per..."
basis, the above Frequency extension applies to each performance after the initial performance.
Exceptions to this Specification are stated in the individual Specifications.
SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. The delay period is only applicable when there is a reasonable expectation the surveillance will be met when performed. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.
If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
SR Applicability 3.0 SEQUOYAH - UNIT 2 3.0-5 Amendment 327, 367 3.0 SR Applicability SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.
This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.
SR Applicability 3.0 SEQUOYAH - UNIT 2 3.0-6 Amendment 367 3.0 SR Applicability Table SR 3.0.2-1 Surveillance Requirement (SR)
Frequency Extension Limit 3.6.12.2 Prior to entering Mode 4 from the SQN Unit 2 Cycle 27 Refueling Outage, but no later than December 1, 2026 3.6.12.3
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 367 RENEWED FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-328
1.0 INTRODUCTION
By application dated February 10, 2025, (Agencywide Documents Access and Management System Accession No. ML25041A086), Tennessee Valley Authority (TVA or the licensee) submitted a license amendment request to the U.S. Nuclear Regulatory Commission (Commission or NRC). TVA requested an amendment to Renewed Facility Operating License DPR-79 for the Sequoyah Nuclear Plant (SQN), Unit 2, in the form of changes to the technical specifications (TSs). The proposed change would revise the SQN, Unit 2, TS Surveillance Requirement (SR) 3.0.2 to extend, on a one-time basis, the surveillance interval for SR 3.6.12.2 (Ice Condenser - Ice Mass Weighing) and SR 3.6.12.3 (Ice Condenser Flow Passage Inspection) that are normally performed on an 18-month frequency in conjunction with a refueling outage. More specifically, the proposed change would extend the due date for SRs 3.6.12.2 and 3.6.12.3 from August 14, 2026, and August 12, 2026, respectively, to prior to entering Mode 4 from the SQN, Unit 2, Cycle 27, Refueling Outage (U2R27), but no later than December 1, 2026.
2.0 REGULATORY EVALUATION
2.1
System Description
SQN, Unit 2, is a Westinghouse-designed pressurized-water reactor with an ice condenser type primary containment. The containment vessel is a welded steel structure made up of a vertical cylindrical wall, a hemispherical dome, and a flat circular base plate encased in concrete. It is divided into three main compartments: (a) the lower compartment, (b) the upper compartment, and (c) the ice condenser compartment. The lower compartment encloses the reactor, steam generators, and associated auxiliary systems equipment. The upper compartment contains the refueling cavity, refueling equipment and polar crane used during refueling and maintenance operations. The ice condenser compartment contains an ice condenser which is an ice bed consisting of borated ice stored in 1944 baskets.
The ice condenser is an annular compartment enclosing approximately 300 degrees of the perimeter of the upper containment compartment, but penetrating the operating deck so that a portion extends into the lower containment compartment. The lower portion has a series of hinged doors exposed to the atmosphere of the lower containment compartment, which, for normal unit operation, are designed to remain closed. At the top of the ice condenser is another set of doors exposed to the atmosphere of the upper compartment, which also remains closed during normal unit operation. Intermediate deck doors, located below the top deck doors, form the floor of a plenum at the upper part of the ice condenser. These doors also remain closed during normal unit operation. The upper plenum area is used to facilitate surveillance and maintenance of the ice bed.
The ice bed consists of a minimum of 1,916,000 pounds (lbs) of ice stored within the ice condenser in 1944 ice baskets. The primary purpose of the ice bed is to provide a large heat sink in the event of release of energy from a design-basis loss-of-coolant accident (LOCA) or main steamline break (MSLB) inside the containment. The ice would absorb energy and limit containment peak pressure and temperature during the accident. Limiting the pressure and temperature reduces the release of fission product radioactivity from containment to the environment in the event of a LOCA or MSLB.
After ice-melt, the containment pressure control is provided by the air return fan system, containment spray train, and residual heat removal spray train. The ice condenser limits the containment pressure below the design pressure for all reactor coolant pipe break sizes up to and including the largest double-ended guillotine break of the reactor coolant system.
2.2 Requested Changes The licensee proposed to revise SR 3.0.2 as follows (underlined text indicates an addition):
The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. In addition, for each of the SRs listed in Table SR 3.0.2-1 the specified Frequency is met if the Surveillance is performed on or before the Frequency extension limit. This extension of the test intervals for these SRs is permitted on a one-time basis.
For Frequencies specified as once, the above interval extension does not apply.
If a Completion Time requires periodic performance on a once per... basis, the above Frequency extension applies to each performance after the initial performance.
Exceptions to this Specification are stated in the individual Specifications.
Additionally, the licensee proposed adding Table SR 3.0.2-1 as follows:
Table SR 3.0.2-1 Surveillance Requirement (SR)
Frequency Extension Limit 3.6.12.2 Prior to entering Mode 4 from SQN Unit 2 Cycle 27 Refueling Outage, but no later than December 1, 2026.
3.6.12.3 2.3 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.
Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. 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) limiting conditions of operation (LCOs); (3) SRs; (4) design features; and (5) administrative controls.
Section 50.36(a)(1) states, that Each applicant for a license authorizing operation of a...
utilization facility shall include in his application proposed technical specifications with the requirements of this section. 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.
As stated in 10 CFR 50.36(b), each license authorizing operation of a production or utilization facility will include TSs. The TSs will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto.
Under 10 CFR 50.36(c)(3), TSs will include SRs, which are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.
The regulation at 10 CFR 50.34(a)(3) cites Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, which establishes the minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the Commission. SQN, Unit 2, was designed to meet the intent of the Proposed General Design Criteria (GDC) for Nuclear Power Plant Construction Permits published in July 1967. However, section 3.1 of the Sequoyah Updated Final Safety Analysis Report (UFSAR) (ML23349A014) addresses the NRC GDC published in July 1971 as Appendix A to 10 CFR Part 50.
The NRC staff considered the following applicable GDCs for the primary containment functional design during its review of this amendment request:
GDC 16, Containment design, which states:
Reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.
GDC 38, Containment heat removal, which states:
A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss-of-coolant accident and maintain them at acceptably low levels.
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.
GDC 39, Inspection of containment heat removal system, which states:
The containment heat removal system shall be designed to permit appropriate periodic inspection of important components, such as the torus, sumps, spray nozzles, and piping to assure the integrity and capability of the system.
GDC 40, Testing of containment heat removal system, which states:
The containment heat removal system 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 active components of the system, and (3) the operability of the system as a whole, and under conditions as close to the 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.
GDC 50, Containment design basis, which states:
The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident. This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and as required by § 50.44 energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergency core cooling functioning, (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters.
3.0 TECHNICAL EVALUATION
The NRC staff evaluated the licensees application to determine whether the proposed changes are consistent with the regulations, guidance, and plant-specific design and licensing basis information discussed in section 2.3 of this safety evaluation.
3.1 Current Surveillance Requirements The 18-month frequency for SR 3.6.12.2 is based on ice storage tests and the allowance built into the required ice mass over and above the mass assumed in the safety analyses. The licensee stated that the minimum weight of 1,145 lbs. of ice per basket in SR 3.6.12.2 contains a 15 percent conservative allowance for ice loss through sublimation, i.e., 15 times higher than the 1 percent sublimation rate for the ice condenser design based on the following statement in SQN UFSAR, Section 6.5.15.3, Sublimation:
For an average temperature of 15°F [Fahrenheit] in the ice condenser compartment, the analytical model predicts a sublimation rate of about 1 percent of the ice mass sublimed per year per ton (12,000 BTU/hr) [British thermal unit/hour] of heat gain to the ice storage compartment.
The minimum weight of 2,225,880 lbs. of ice in the ice condenser also contains an additional 1 percent conservative allowance to account for systematic error in weighing instruments. The licensees operating experience with performing SR 3.6.12.2 has verified that, with the 18-month SR frequency, the weight requirements are maintained with no significant degradation between surveillances.
SR 3.6.12.3 verifies by visual inspection that the accumulation of ice on structural members comprising flow channels through the ice bed is 15 percent blockage of the total flow area through the ice condenser. The maximum 15 percent blockage of the flow area is based on the analysis described in SQN UFSAR, section 6.2.1.3.3, under the title, Initial Pressure Peaks (Assuming a 15% reduction in flow area through the ice condenser).
The next due dates for SR 3.6.12.2 and SR 3.6.12.3 are August 14, 2026, and August 12, 2026, respectively. These dates are 18 months from the date the last surveillance was initiated including a 25-percent extension of the date as allowed by SR 3.0.2.
3.2 Evaluation of Changes to Technical Specifications The NRC staff reviewed the acceptability of the proposed changes to the TSs by evaluating whether, among other things, the changes provide reasonable assurance that public health and safety will be protected. The NRC staff also verified that the proposed changes to the TSs will continue to ensure that the LCO will be met.
The proposed change would extend the due date for SR 3.6.12.2 and SR 3.6.12.3 prior to entering Mode 4 from U2R27, but no later than December 1, 2026.
The licensee used NRC guidance in Generic Letter (GL) 91-04, Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle (ML031140501),
to evaluate the proposed one-time surveillance extensions. GL 91-04 provides guidance for evaluating the impact of adopting a 24-month surveillance test interval. The licensees evaluation consisted of performing history reviews for determining ice sublimation rates. The licensee stated that for SQN, Unit 1, Cycle 25, the average sublimation rate was found to be 3.97 percent, and for SQN, Unit 2, Cycle 26, the average sublimation rate was 4.17 percent.
Therefore, as noted in section 3.1 above, the 15 percent sublimation allowance included in the minimum weight of 1,145 lbs. of ice per basket is conservative. At the 15 percent per cycle sublimation rate, the TS 3.6.12 minimum ice mass is intended to provide sufficient ice to last for a typical 18-month fuel cycle.
The SQN, Unit 2, Cycle 28 is currently planned to commence in November 2026, which is approximately 26 months from the last ice weighing surveillance that was initiated on September 26, 2024. To calculate the sublimation during the proposed extended 26 months, the licensee assumed a linear sublimation rate of ice as given below.
Sublimation (extended cycle) = Sublimation (typical cycle) x (time (extended cycle) /
time (typical cycle))
Sublimation (extended cycle) = 15% x ((26 months) / (18 months))
Sublimation (extended cycle) = 21.67% (rounded)
The licensee stated that for conservatism, the initial SQN, Unit 2, Cycle 27, ice bed was loaded with 2,631,962 lbs. of ice instead of the TS 3.6.12 required minimum weight of 2,225,880 lbs.
Applying the calculated 21.67 percent sublimation to the loaded ice weight of 2,631,962 lbs., at the end of the extended operating cycle, the ice bed would still contain approximately 2,061,704 lbs. of ice which is approximately 145,704 lbs. above the analytical limit of 1,916,000 lbs.
required for the mitigation of a design-basis LOCA or MSLB.
The NRC staff finds the proposed increase in the surveillance interval for SR 3.6.12.2 acceptable because there is margin in the remaining ice mass after sublimation for the mitigation of design-basis LOCA or MSLB during SQN, Unit 2, Cycle 27.
As noted in section 3.1 above, SR 3.6.12.3 verifies by visual inspection that the accumulation of ice on structural members comprising flow channels through the ice bed is 15 percent blockage of the total flow area through the ice condenser. This SR ensures that the analysis performed for the sub-compartment response to a design-basis LOCA with partial blockage of the ice condenser flow channels remains valid (see SQN UFSAR section 6.2.1.3.3).
The licensee proposed to extend SR 3.6.12.3 until prior to entering Mode 4 from U2R27, currently scheduled to be conducted in fall 2026, but not later than December 1, 2026. The current frequency of SR 3.6.12.3 is 18 months plus an allowable 25 percent extension per SR 3.0.2, which states, The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. As noted in the LAR, SR 3.6.12.3 was last initiated on September 24, 2024, and successfully completed on October 9, 2024, which means the next surveillance is due on or before August 12, 2026.
In its LAR, the licensee stated that operating experience has demonstrated that the ice bed is the most restrictive flow region because of the accumulation of ice on the lattice frames and panel walls. The LAR also stated that there is no mechanistically credible method for ice to accumulate in the passages during plant operation. The LAR further stated that ice cannot accumulate in the flow passages during plant operation.
The licensee has implemented improved practices that were approved in TSTF-336, Revision 1, Ice bed flow channel blockage surveillance requirement (ML040630117), which revised the SR to clarify the requirements and extend the frequency of the inspections required by SR 3.6.12.3.
TSTF-336 states that the SR is performed at the end of each refueling outage and that this ensures that the flow passages will remain open for the remainder of the operating cycle. The performance of the SR is to ensure that maintenance activities or changes during outages do not cause blockage of the flow passages. Because the last performance of the SR successfully observed that there is sufficient open flow area, and there is no credible means for ice to accumulate in the flow passages during operation, the relatively short extension of the SR due date should not result in ice accumulation that will block the flow passages. The NRC staff concludes that the last successful performance of the SR provides adequate assurance that sufficient flow area through the ice bed will be maintained for the additional time.
The NRC staff finds the proposed increase in the surveillance interval for SR 3.6.12.3 acceptable because there is no credible mechanism for ice to form and block passages during operation and the licensee successfully performed the SR during its last outage, thus ensuring that the flow passages were not blocked.
3.3 Technical Conclusion Based on its review of the licensees request, the NRC staff finds that there will be sufficient ice mass available at the end of the requested extension of the SRs. In addition, satisfactory completion of SR 3.6.12.3 that verified the level of obstruction of flow passages through the ice bed is acceptable and the observation that there is no credible method for ice to accumulate in the passages during operation provides reasonable assurance that the ice condenser flow paths remain free from unanalyzed obstructions. Therefore, the NRC staff finds the licensees request to modify TS 3.0.2 to insert Table SR 3.0.2-1 to allow for a one-time extension of SRs 3.6.12.2 and 3.6.12.3 intervals to be acceptable. The insertion of the table has no effect on the interpretation of SR 3.0.2 other than to allow these two SR frequencies to be extended.
The only changes to the TS are to the SR frequencies described in this safety evaluation. As required by 10 CFR 50.36(c)(3), SRs are required to be included in TSs. The requirements of 10 CFR 50.36(c)(3) will be met because the evaluation demonstrates that the subject SRs will continue to assure that the necessary quality of system and components is maintained, that facility operation will be within safety limits, and that the LCO will be met during the extended interval. Therefore, the NRC staff concludes that 10 CFR 50.36 requirements will continue to be met.
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 18, 2025. 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 surveillance requirements. 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 previously issued a proposed finding that the amendment involves no significant hazards consideration published in the Federal Register on April 15, 2025 (90 FR 15727), 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: A. Sallman, NRR S. Smith, NRR Date: September 22, 2025
SUBJECT:
SEQUOYAH NUCLEAR PLANT, UNIT 2 - ISSUANCE OF AMENDMENT NO. 367 REGARDING REVISION OF TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT 3.0.2 TO EXTEND ICE MASS SURVEILLANCE REQUIREMENTS ON ONE-TIME BASIS (EPID L-2025-LLA-0032) DATED SEPTEMBER 22, 2025 DISTRIBUTION:
PUBLIC RidsNrrDorlLpl2-2 RidsNrrPMSequoyah RidsNrrLAABaxter RidsACRS_MailCTR RidsNrrDssSnsb RidsNrrDssStsb RidsRgn2MailCenter ASallman, NRR SSmith (DSS), NRR ADAMS Accession No.: ML25252A202 NRR-058 OFFICE NRR/DORL/LPLII-2/PM NRR/DORL/LPLII-2/LA NRR/DSS/SNSB/BC NAME KGreen ABaxter NDiFrancesco DATE 09/09/2025 09/10/2025 07/10/2025 OFFICE NRR/DSS/STSB/BC NRR/DORL/LPLII-2/BC NRR/DORL/LPLII-2/PM NAME SMehta DWrona KGreen DATE 09/19/2025 09/22/2025 09/22/2025