ML21084A190

Issuance of Amendment Nos. 355 and 348 Regarding Revision to Technical Specification Table 3.3.3-1, Post Accident Monitoring Instrumentation

ML21084A190
Person / Time
Site:	Sequoyah
Issue date:	05/04/2021
From:	Michael Wentzel Plant Licensing Branch II
To:	Jim Barstow Tennessee Valley Authority
Wentzel M
References
EPID L 2020 LLA 0132
Download: ML21084A190 (21)

Text

May 4, 2021 Mr. James Barstow Vice President, Nuclear Regulatory Affairs and Support Services Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801

SUBJECT:

SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 355 AND 348 REGARDING REVISION TO TECHNICAL SPECIFICATION TABLE 3.3.3-1, POST ACCIDENT MONITORING INSTRUMENTATION (EPID L-2020-LLA-0132)

Dear Mr. Barstow:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 355 to Renewed Facility Operating License No. DPR-77, and Amendment No. 348 to Renewed Facility Operating License No. DPR-79, for the Sequoyah Nuclear Plant, Units 1 and 2, respectively. These amendments are in response to your application dated June 16, 2020.

The amendments revise Technical Specification Table 3.3.3-1, Post Accident Monitoring Instrumentation, required actions and completion times for Functions 15 a, b, and c, Reactor Vessel Level Instrumentation. Additionally, the amendments delete Note g from Table 3.3.3-1, Function 15.c from the Unit 2 Technical Specifications and remove License Condition 2.C.(26) from the Unit 2 Renewed Facility Operating License. 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/

Michael J. Wentzel, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-327 and 50-328

Enclosures:

1. Amendment No. 355 to DPR-77

2. Amendment No. 348 to DPR-79

3. Safety Evaluation cc: Listserv

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 355 Renewed License No. DPR-77

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Tennessee Valley Authority (the licensee) dated June 16, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (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.

Enclosure 1

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-77 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 355 are hereby incorporated in 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 from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by David J.

David J. Wrona Date: 2021.05.04 13:30:38 Wrona -04'00' David J. 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: May 4, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 355 SEQUOYAH NUCLEAR PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Replace page 3 of the Renewed Facility Operating License with the attached page 3.

Replace the following page of the Appendix A Technical Specifications with the attached page.

The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 3.3.3-5 3.3.3-5

(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 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 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. 355 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 Authoritys 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 TVAs Final Safety Analysis Report as amended as being essential;

b. Modification of test objectives, methods, or acceptance criteria for any test identified in Section 14 of TVAs Final Safety Analysis Report as amended as being essential; Amendment No. 355 Renewed License No. DPR-77

PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 2 of 2)

Post Accident Monitoring Instrumentation CONDITION REFERENCED FROM REQUIRED FUNCTION REQUIRED CHANNELS ACTION G.1

15. Reactor Vessel Level Instrumentation

a. Dynamic Range 2 I

b. Lower Range 2 I

c. Upper Range 2 I

16. Containment Area Radiation Monitors

a. Upper Compartment 1 I

b. Lower Compartment 1 I

17. Neutron Flux

a. Source Range 2(c) H

b. Intermediate Range 2 H

18. ERCW to AFW Valve Position

a. Motor Driven Pumps 2(d) H

b. Turbine Driven Pump 2(d) H

19. Containment Isolation Valve Position 2 per penetration H flowpath(e)(f)

(c) Source Range outputs may be disabled above the P-6 (Block of Source Range Reactor Trip) setpoint.

(d) A channel consists of two valve position indicators associated with the in-series valves in a single suction line.

(e) Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

(f) Only one position indication channel is required for penetration flow paths with only one installed control room indication channel.

SEQUOYAH - UNIT 1 3.3.3-5 Amendment 334, 355

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 348 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 June 16, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 2

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. 348 are hereby incorporated in 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 from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by David David J. J. Wrona Date: 2021.05.04 Wrona 13:31:13 -04'00' David J. 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: May 4, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 348 SEQUOYAH NUCLEAR PLANT, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 Replace pages 3 and 13a of the Renewed Facility Operating License with the attached pages 3 and 13a.

Replace the following page of the Appendix A Technical Specifications with the attached page.

The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 3.3.3-5 3.3.3-5

(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. 348 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. 348 Renewed License No. DPR-79

- 13a -

relocation of the requirements to the specified documents, as described in Table R, Relocated Specifications and Removed Detail Changes, attached to the NRC staff's Safety Evaluation, which is enclosed in this amendment.

2. Schedule for New and Revised Surveillance Requirements (SRs) The schedule for performing SRs that are new or revised in License Amendment 327 shall be as follows:

(a) For SRs that are new in this amendment, the first performance is due at the end of the first Surveillance interval, which begins on the date of implementation of this amendment.

(b) For SRs that existed prior to this amendment, whose intervals of performance are being reduced, the first reduced Surveillance interval begins upon completion of the first Surveillance performed after implementation of this amendment.

(c) For SRs that existed prior to this amendment, whose intervals of performance are being extended, the first extended Surveillance interval begins upon completion of the last Surveillance performed prior to implementation of this amendment.

(d) For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance subject to the modified acceptance criteria is due at the end of the first Surveillance interval that began on the date the Surveillance was last performed prior to the implementation of this amendment.

(26) DELETED (27) Adoption of 10 CFR 50.69, "Risk-Informed categorization and treatment of structures, systems and components for nuclear power plants" (1) TVA is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) model to evaluate risk associated with internal events, including internal flooding; using the firesafe shutdown equipment list in the SQN Fire Protection Report referenced in the Updated Final Safety Analysis Report to evaluate internal fire events; the NUMARC 96-01 shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non-PRA evaluations that are based on the Amendment No. 338, 340, 348 Renewed License No. DPR 79

PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 2 of 2)

Post Accident Monitoring Instrumentation CONDITION REFERENCED FROM REQUIRED FUNCTION REQUIRED CHANNELS ACTION G.1

15. Reactor Vessel Level Instrumentation

a. Dynamic Range 2 I

b. Lower Range 2 I

c. Upper Range 2 I

16. Containment Area Radiation Monitors

a. Upper Compartment 1 I

b. Lower Compartment 1 I

17. Neutron Flux

a. Source Range 2(c) H

b. Intermediate Range 2 H

18. ERCW to AFW Valve Position

a. Motor Driven Pumps 2(d) H

b. Turbine Driven Pump 2(d) H

19. Containment Isolation Valve Position 2 per penetration H flowpath(e)(f)

(c) Source Range outputs may be disabled above the P-6 (Block of Source Range Reactor Trip) setpoint.

(d) A channel consists of two valve position indicators associated with the in-series valves in a single suction line.

(e) Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

(f) Only one position indication channel is required for penetration flow paths with only one installed control room indication channel.

SEQUOYAH - UNIT 2 3.3.3-5 Amendment 327, 338, 348

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 355 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-77 AND AMENDMENT NO. 348 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328

1.0 INTRODUCTION

By application dated June 16, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20169A497), the Tennessee Valley Authority (TVA; the licensee) submitted a license amendment request (LAR) for Sequoyah Nuclear Plant (Sequoyah), Units 1 and 2. The proposed amendments would revise the Sequoyah, Units 1 and 2, Technical Specification (TS) Table 3.3.3-1, Post Accident Monitoring Instrumentation, required actions and completion times for Functions 15 a, b, and c, Reactor Vessel Level Instrumentation. Additionally, the amendments would delete Note g from Unit 2 Table 3.3.3-1, Function 15.c and would remove License Condition 2.C.(26) from the Unit 2 Renewed Facility Operating License.

2.0 REGULATORY EVALUATION

2.1 Reactor Vessel Level Instrumentation System Instrumentation Design The Reactor Vessel Level Instrumentation System (RVLIS) consists of the reactor vessel dynamic range channels, lower range level channels, and upper range level channels. The RVLIS performs the following functions: 1) the dynamic range channels indicate the formation of voiding in the RCS during forced flow conditions; 2) the lower range channels detect the approach to inadequate core cooling (ICC) for beyond design-basis events; and 3) the upper range channels indicate the presence and measure the size of steam void or non-condensable gas bubble in the reactor vessel during natural circulation conditions in the reactor coolant system (RCS).

TS 3.3.3, Post Accident Monitoring (PAM) Instrumentation, Limiting Condition for Operation 3.3.3, requires PAM instrumentation for each Function in Table 3.3.3-1 to be Enclosure 3

operable. Table 3.3.3-1 requires two channels of the RVLIS instruments to be operable for Sequoyah, Units 1 and 2. Each channel employs differential pressure transmitters to measure the pressure drop from the bottom of the reactor vessel to the hot legs and from the hot legs to the top of the vessel. These transmitters have the following three different ranges to cover different level ranges and plant conditions:

1. The reactor vessel dynamic range channels, TS Table 3.3.3-1 Function 15.a, provides reactor vessel level indication from 0 percent to 120 percent. This range provides an indication of reactor core and internals pressure drop for any combination of operating reactor coolant pumps and monitors coolant conditions and indicates formation of voiding in the RCS during forced flow conditions.

2. The reactor vessel lower range channels, TS Table 3.3.3-1 Function 15.b, provides reactor vessel level indication from 0 percent to 70 percent, which represents the reactor vessel level from the bottom of the reactor to the center of the hot legs. This indication is only used in emergency operating procedures to monitor the adequacy of core cooling based on reactor water-level when reactor coolant pumps are stopped.

3. The reactor vessel upper range channels, TS Table 3.3.3-1 Function 15.c, provides reactor vessel level indication from 64 percent to 104 percent, which represents the reactor vessel level from the center of the hot leg pipes to the top of the reactor vessel head. The upper range channels will indicate the presence and measure a size of steam void or non-condensable gas bubble in the reactor vessel during natural circulation conditions in the RCS.

2.2 Proposed Changes In July 2019, the licensee submitted an exigent, one-time LAR to avoid an unplanned shutdown of Sequoyah, Unit 2, during Cycle 23 due to two inoperable channels of the RVLIS upper range.

The U.S. Nuclear Regulatory Commission (NRC) approved this request by letter dated July 18, 2019 (ADAMS Accession No. ML19196A221). The licensee submitted this LAR to avoid future plant shutdowns due to potential RVLIS channel inoperability.

The licensee proposed to modify TS 3.3.3, Table 3.3.3-1, as indicated below (deleted text indicated by strikethrough, added text indicated by underline):

CONDITION REFERENCED REQUIRED FROM REQUIRED FUNCTION CHANNELS ACTION G.1

15. Reactor Vessel Level Instrumentation

a. Dynamic Range 2 HI

b. Lower Range 2 HI

c. Upper Range 2 HI

Condition I directs the licensee to take action in accordance with TS 5.6.5, Post Accident Monitoring Report. TS 5.6.5 states that [w]hen a report is required by Condition B or I of LCO 3.3.3, Post Accident Monitoring (PAM) Instrumentation, a report shall be submitted [to the NRC] within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. Accordingly, this proposed TS would allow the establishment of alternate means of verifying post-accident core cooling with the RVLIS instruments inoperable, rather than requiring a plant shutdown.

Regarding the use of alternate post-accident monitoring instrumentation, the licensee identified the following TS 3.3.3 instrumentation that could be utilized to provide diverse information for verifying existence of adequate inventory for core cooling, should one or more of the three RVLIS Functions become inoperable.

1. Pressurizer Level - Wide Range, TS Table 3.3.3-1 Function 7, monitors RCS inventory to confirm that the plant is in a safe shutdown condition.

2. RCS Subcooling instrumentation, TS Table 3.3.3-1 Function 12, is used to determine safety injection termination or re-initiation and maintenance of subcooling during depressurization. The channels provide indication over a range of 200 degrees Fahrenheit subcooled to 35 degrees Fahrenheit superheat.

3. Core-exit thermocouples (CETs), TS Table 3.3.3-1 Functions 14.a, 14.b, 14.c, and 14.d, (designated as In-core Thermocouples), provide the information for use to verify that the core is being adequately cooled, verify that RCS remains subcooled, and for monitoring the potential for fuel clad breach. The channels provide indication over a range of 0 degrees Fahrenheit to 2300 degrees Fahrenheit.

For Sequoyah, Unit 2, the licensee proposed to delete Note (g) from Table 3.3.3-1 and License Condition 2.C.(26), both of which were added as a part of the July 2019 exigent amendment and were only applicable for the remainder of Cycle 23, which has since ended.

Note (g)

(g) The Upper Range Reactor Vessel Level Instrumentation is not required to be operable for the remainder of Cycle 23. If SQN Unit 2 enters Mode 5 prior to the Unit 2 Cycle 23 refueling outage, TVA will further validate the cause of the inoperability of the Upper Range Reactor Vessel Level Instrumentation and the Upper Range Reactor Vessel Level Instrumentation will be restored to OPERABLE status prior to plant startup.

Regardless of the above action, the Upper Range Reactor Vessel Level Instrumentation will be restored to OPERABLE status no later than the end of the Unit 2 Cycle 23 refueling outage.

License Condition 2.C.(26)

TVA will implement the compensatory measures described in Section 3.8, "Additional Compensatory Measures," of TVA letter CNL-19-072, dated July 14, 2019, during the

timeframe the Upper Range Reactor Vessel Level Instrumentation is not required to be operable for the remainder of Cycle 23. If the Upper Range Reactor Vessel Level Instrumentation is returned to operable status prior to the end of Cycle 23, then these compensatory measures are no longer required.

2.3 Regulatory Requirements Section 50.36, Technical specifications, of Title 10 of the Code of Federal Regulations (10 CFR) establishes the regulatory requirements related to the content of TSs.

Paragraph 50.36(a)(1) requires an application for an operating license to include proposed TSs.

A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the TSs.

Pursuant to 10 CFR 50.36, TSs for operating reactors are required, in part, to include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements; (4) design features; and (5) administrative controls.

Paragraph 50.36(c)(2) of 10 CFR states that LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility, and when an LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.

Appendix A, General Design Criteria (GDC) for Nuclear Power Plants, to 10 CFR Part 50 establishes the minimum requirements for the principal design criteria for water-cooled nuclear power plants. The following GDC is applicable for this review:

Criterion 13 - Instrumentation and control.

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

Section 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, of 10 CFR requires, in part, that the emergency core cooling system be designed with sufficient margin to assure that the design safety limits specified in 10 CFR 50.46(b) are met during loss-of-coolant accidents.

Regulatory Guide (RG) 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, Revision 5 (ADAMS Accession No. ML18136A762), describes a method that is acceptable to the NRC staff for use in complying with the agencys regulations with respect to satisfying criteria for accident monitoring instrumentation in nuclear power plants.

NUREG-0737, Clarification of TMI [Three Mile Island] Action Plan Requirements (ADAMS Accession No. ML051400209), incorporated all TMI-related items approved for implementation by the NRC. Enclosures 1 and 2 in NUREG-0737 contain an itemized listing of requirements.

TMI action Item II.F.2 is applicable to this review. Item II.F.2 required licensees to provide a description of any additional instrumentation or controls (primary and backup) proposed for the plants to supplement existing instrumentation (including primary coolant saturation monitors) in order to provide unambiguous, easy-to-interpret indication of ICC. The indication of ICC must be unambiguous in that it should have the following properties:

It must indicate the existence of ICC caused by various phenomena (i.e., high-void fraction-pumped flow as well as stagnant boil-off); and It must not erroneously indicate ICC because of the presence of an unrelated phenomenon.

The indication must give advanced warning of the approach of ICC. The indication must cover the full-range from normal operation to complete core uncovery. For example, water-level instrumentation may be chosen to provide advanced warning of two-phase level drop to the top of the core and could be supplemented by other indicators such as in-core and CETs provided that the indicated temperatures can be correlated to provide indication of the existence of ICC and to infer the extent of core uncovery. Alternatively, full-range level instrumentation to the bottom of the core may be employed in conjunction with other diverse indicators such as CETs to preclude misinterpretation due to any inherent deficiencies or inaccuracies in the measurement system selected.

NUREG-1431, Revision 4, Standard Technical Specifications [STSs], Westinghouse Plants (ADAMS Accession No. ML12100A222), is available for licensees to use to revise or upgrade their technical specifications consistent with the requirements in 10 CFR 50.36.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the licensees application to determine if the proposed changes are consistent with the guidance, regulations, and plant-specific design and licensing basis information discussed in Section 2.3 of this safety evaluation.

3.1 Evaluation of the Use of Alternate Indicators The licensees proposed change to Table 3.3.3-1, Functions 15.a, 15.b, and 15.c, to revise the Condition referenced from H to I (evaluated in Section 3.2 of this safety evaluation) would allow for the use of alternative indicators should the operability requirements of Table 3.3.3-1 not be met. Sections 3.3.1 through 3.3.3 of the LAR describe the specific functions of the RVLIS dynamic range, lower range, and upper range, respectively, and identifies other PAM instruments (such as CETs, Pressurizer Level, and RCS Subcooling Margin instruments) that are available as alternative methods. The NRC staffs evaluation of the use of alternative indicators is below.

3.1.1 Compliance with the Requirements of GDC 13 and 10 CFR 50.46 The NRC staff confirmed that RVLIS is a Type B, Category 1, RG 1.97 PAM function that provides indication only. The RVLIS indication is not used by the main control room operators to perform manual actions required for safety systems to accomplish their functions for the Updated Final Safety Analysis Report (UFSAR) Chapter 15 accident analysis.

Section 3.2 of the LAR identifies alternate indicators that could be used in the event that one or more RVLIS function is unavailable. All of the alternate indications are required to be operable per TS 3.3.3 and provide the capability to monitor for ICC. The NRC staff determined that the alternate indications provided by CETs, Pressurizer Level, and RCS Subcooling Margin instruments will provide an acceptable means for verifying the existence of adequate core cooling when the channels of the RVLIS dynamic range, lower range, and upper range instruments are inoperable. The NRC staff reviewed the information provided in the LAR and concludes that alternate indications are available and adequate for each of the three RVLIS Functions: Function 15.a, which relates to indication of the formation of voiding in the RCS during forced flow conditions; Function 15.b, which relates to detection of the approach to ICC for beyond design-basis events; and Function 15.c, which relates to indication of the presence and size of a steam void or non-condensable gas bubble in the reactor vessel during natural circulation conditions in the RCS.

Regarding the potential for the use of alternative indicators to impact UFSAR Chapter 15 accident analyses, the licensee indicated in Item 2 of Section 3.3.2 of the LAR that unexpected variations in RVLIS lower range indication is one of the key indications for diagnosing a long-term degraded core cooling condition following a large-break LOCA (LBLOCA) design-basis event. Section 3.3.2 of the LAR also states that based on Westinghouse analyses, certain break locations could produce long-term degraded cooling due to inadequate venting of steam produced in the core. The LAR further states that, for those LBLOCAs, CET indication rising unexpectedly several hours after a LBLOCA is an alternate indication that could be used to diagnose a long-term degraded cooling condition and initiate mitigating actions. The NRC staff concludes that, given that the probability of a LBLOCA is extremely low, the risk of a degraded core cooling condition is small, as it could occur only during LBLOCAs at certain break locations. Further, the NRC staff concludes that the CET would provide sufficient indication of ICC. Therefore, the NRC staff finds that during a loss of RVLIS lower range channels, the use of CET indication provides reasonable assurance that the analysis of record for the LOCA during long-term core cooling would remain valid.

Based on the above, the NRC staff finds that for plant conditions with inoperable RVLIS channels, the Updated Final Safety Analysis Report Chapter 15 accident analysis remains valid.

Therefore, the NRC staff finds that the licensees proposed use of alternative indications will continue to meet the requirements of GDC 13 regarding the provision of instrumentation and control for monitoring reactor conditions and 10 CFR 50.46 regarding the acceptable criteria for the emergency core cooling system performance and is acceptable.

3.1.2 Compliance with the Guidance of TMI Action Item II.F.2 in NUREG-0737 Related to Instrumentation for Detection of the Approach to ICC Section 3.3.2 of the LAR describes that RVLIS lower range provides indications to detect the approach to and recovery from ICC in accordance with the guidance of TMI Action Item II.F.2 in NUREG-0737. Section 3.3.2 of the LAR also describes the alternative means of monitoring as being indicated by CET when the RVLIS lower range channels are inoperable. Other additional PAM instruments (Pressurizer Level and RCS Subcooling Margin) provide diverse information for verifying adequate inventory for core cooling.

The NRC staff finds that the licensees proposed use of alternate PAM instrumentation to perform the Function of the RVLIS lower range meets the guidance of NUREG-0737 II.F.2 to provide instrumentation for detection of the approach to ICC. NUREG-0737 II.F.2 does not apply to other RVLIS Functions.

3.2 Evaluation of TS Changes and Deletion of License Condition 2.C.(26) 3.2.1 Change to Reference Condition I for Function 15 The LCO for TS 3.3.3 requires that the PAM instrumentation for each function in Table 3.3.3-1 be operable in Modes 1, 2, and 3. Required Action (RA) C in the Actions Table requires that, with one or more functions with two required channels inoperable, one channel must be restored to operable status within 7 days. RA G.1 requires that for the situation with the required action and completion time of Conditions C, D, E, or F not met, the condition referenced in Table 3.3.3-1 for the associated affected instrument must be entered. Table 3.3.3-1 contains a Condition Referenced from Required Action G.1 column that requires entry into Condition H for RVLIS. Condition H requires the plant to be in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The proposed change would revise the Condition referenced from H to I for Functions 15.a, 15.b, and 15.c. Condition I, which requires that action be taken per TS 5.6.5, directs the licensee to submit a report to the NRC within the following 14 days outlining the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the operability of the instrumentation channels.

The NRC staff notes that Condition H would still be referenced for Function 7 (Wide Range Pressurizer Level instrument), Function 12 (RCS Subcooling Margin monitor) and Function 14 (CETs). Condition H for Functions 7, 12 and 14 requires the plant to be in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, for situations where three required channels of the Pressurizer Level instrument (Function 7) are inoperable, or two channels of the RCS Subcooling Margin monitor (Function 12) or CETs (Function 14) are inoperable during Modes 1 and 2. Those requirements assure that in the event that the RVLIS instruments (Function 15.a, 15.b, and 15.c) are inoperable during Modes 1 through 3, the alternate means (including Wide Range Pressurizer Level, RCS Subcooling Margin monitor, and CETS) remain available to provide information for verifying existence of adequate inventory for core cooling and meet the intent of the TMI Action Item II.F.2 guidance.

The NRC staff also compared the proposed changes to the corresponding standard TS in NUREG-1431 and found that for the similar situation, the proposed changes for Sequoyah, Units 1 and 2, are consistent with STS 3.3.3 in NUREG-1431.

Because the proposed changes 1) do not change the Updated Final Safety Analysis Report Chapter 15 analysis (discussed in Section 3.1.1 above), 2) do not change licensee commitments for meeting the intent of the guidance of TMI Action Item II.F.2 (discussed in Section 3.1.2 above), and 3) are consistent with the standard TS in NUREG-1431, the NRC staff finds that the proposed TS changes are acceptable.

3.2.2 Deletion of Note (g) in Table 3.3.3-1 for Function 15.c and License Condition 2.C.(26) for Sequoyah, Unit 2 The licensee proposed to delete Note (g) in TS Table 3.3.3-1 and License Condition 2.C.(26) for Sequoyah, Unit 2. Note (g) and License Condition 2.C.(26) were added as conditions of approval of the licensees July 2019 exigent LAR and were valid only during Cycle 23, which has since ended and RVLIS upper range operability has been restored. The NRC staff finds that Note (g) and License Condition 2.C.(26) are both obsolete, therefore, the proposed change is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Tennessee State official was notified of the proposed issuance of the amendments on March 25, 2021. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. 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 has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on August 11, 2020 (85 FR 48572). 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: S. Sun, NRR C. Cheung, NRR Dated: May 4, 2021

ML21084A190 *by e-mail OFFICE NRR/DORL/LPLII-2/PM NRR/DORL/LPLII-2/LA NRR/DSS/SNSB/BC NAME MWentzel RButler SKrepel DATE 04/06/2021 04/05/2021 02/08/2021 OFFICE NRR/DE/EICB/BC* NRR/DSS/STSB/BC OGC - NLO*

NAME MWaters VCusumano MYoung DATE 03/18/2021 03/31/2021 04/20/2021 OFFICE NRR/DORL/LPLII-2/BC NRR/DORL/LPLII-2/PM NAME DWrona MWentzel DATE 05/04/2021 05/04/2021