ML25182A175

CFR 50.59 and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report; and Commitment Summary Report

ML25182A175
Person / Time
Site:	Sequoyah
Issue date:	06/30/2025
From:	Michael K Tennessee Valley Authority
To:	Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation, Document Control Desk
References
Download: ML25182A175 (1)

Text

Sequoyah Nuclear Plant, Post Office Box 2000, Soddy Daisy, Tennessee 37384 June 30, 2025 10 CFR 50.59 10 CFR 72.48 10 CFR 50.71 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327, 50-328, and 72-034

Subject:

10 CFR 50.59 and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report; and Commitment Summary Report

Reference:

TVA letter to NRC, 10 CFR 50.59 and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report; Commitment Summary Report; and Update to Fire Protection Report, dated October 11, 2023 In accordance with 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2), Enclosure 1 is the Sequoyah Nuclear Plant (SQN), Units 1 and 2, Summary Report regarding the implemented changes, tests, and experiments for which evaluations were performed in accordance with 10 CFR 50.59(c) and 10 CFR 72.48(c). The summarized evaluations provided in the enclosure were implemented since the Reference Letter through April 14, 2025.

Since last reported in the Reference Letter, SQN has revised a regulatory commitment in accordance with NEI 99-04, the Nuclear Energy Institute's "Guidelines for Managing NRC

[Nuclear Regulatory Commission] Commitment Changes," as endorsed in NRC Regulatory Issue Summary 2000-17. The commitment change summary is provided in Enclosure 2.

U.S. Nuclear Regulatory Commission Page 2 June 30, 2025 There are no new commitments contained in this letter. If you have any questions concerning this submittal, please contact Mr. Rick Medina, SQN Compliance Manager at (423) 843-8129 or rmedina4@tva.gov.

Respectfully, "n /)

~fvl.}-vl,.JL.x' Kevin M. Michael Site Vice President Sequoyah Nuclear Plant Enclosures 1.

10 CFR 50.59, and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report

2.

Commitment Change Report cc (w/Enclosures ):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Sequoyah Nuclear Plant Director, Division of Fuel Management, Office of Nuclear Material Safety and Safeguards NRC Project Manager - Sequoyah Nuclear Plant

E1-2 ENCLOSURE 1 SEQUOYAH NUCLEAR PLANT 10 CFR 50.59 AND 10 CFR 72.48 CHANGES, TESTS, AND EXPERIMENTS

SUMMARY

REPORT DESIGN CHANGES DESCRIPTION SAFETY ANALYSIS SQN-18-190 This modification addresses Appendix R Ground Fault Equivalent Hot Short (GFEHS) issues associated with the Reactor Coolant System Head Vents (RCSHVs) specifically, the reactor head vent throttle valves and reactor head vent isolation valves. The following aspect of the activity has been identified in 10 CFR 50.59 screening review as having an adverse effect on Updated Final Safety Analysis Report (UFSAR) design functions and is addressed in this 10 CFR 50.59 Evaluation: Addition of DC/DC converters for the head vent throttle valve positioners.

The new head vent throttle valve positioner A-A for the Unit 1 train A throttle valve will be installed in Unit 2 side of the auxiliary building. Due to design differences between the existing valve positioner and the replacement valve positioner, a new DC/DC converter is required to provide Appendix R separation between the 125VDC station power and the 125VDC power used to control the throttle valve. This will require installation of new cables.

A new Unit 1 head vent throttle valve positioner A-B will replace the existing positioner. Due to design differences between the existing valve positioner and the replacement valve positioner, a DC/DC converter is required to maintain the required Appendix R separation. Thus, this design installs a replacement valve positioner and associated DC/DC converter.

Based on the responses to all 10 CFR 50.59 Evaluation Questions it is concluded that this activity to install new DC/DC converters for head vent throttle valve positioners:

Does not result in increase of frequency and consequences of an accident previously evaluated in UFSAR or create a possibility for an accident of a different type than any previously evaluated in the UFSAR.

Does not result in increase in the likelihood of occurrence and consequences of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the Final Safety Analysis Report as updated (UFSAR) or create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in UFSAR.

Does not affect design basis limit for a fission product barrier as described in the UFSAR.

Does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

The new DC/DC converters are safety related (SR) components whose reliability is equal to the reliability of the components in the 125VDC power supply circuit to the positioners. The new DC/DC converters are compatible with positioner power requirements. The DC/DC converters, similarly to the positioners, are analog devices and therefore, this activity does not involve analog to digital conversion. Operation of the RCSHVs due to this activity is not altered. The RCSHVs will respond the same way to any event described in

E1-3 The existing cable from 125VDC vital battery board II will be disconnected and spared and new cables will be installed.

The new head vent throttle valve positioner A-A for the Unit 2 train A valve will be installed in Unit 1 side of the auxiliary building. Due to design differences between the existing valve positioner and the replacement valve positioner, a new DC/DC converter is required to provide Appendix R separation between the 125VDC station power and the 125VDC power used to control the throttle valve. To support the new valve positioner two new instrumentation cables will be installed and will connect to previously spared Unit 1 cables that will be re-labelled to Unit 2. A new cable will be installed from the DC/DC converter output to the new valve positioner.

A new Unit 2 head vent throttle valve positioner A-B will replace existing positioner. Due to design differences between the existing valve positioner and the replacement valve positioner, a DC/DC converter is required to maintain the required Appendix R separation. Thus, this design installs a replacement valve positioner and associated DC/DC converter.

The existing cable will be spared, and new cables will be installed.

UFSAR. The cables from DC/DC converters to the positioners comply with station requirements, such as routing, seismic, load and voltage drop. Based on above statements, there is no more than a minimal increase in the likelihood of occurrence of a malfunction of the affected SSCs and there is no increase in the consequences of an accident previously evaluated in the UFSAR and, therefore, there is no impact on any of the SSCs functions.

Therefore, this activity may be implemented per plant procedures and does not require a license amendment request.

SQN-20-1620 Sequoyah Nuclear Plant (SQN) has four Westinghouse Veritrak 75HC-3000 hand indicating controllers (HICs) which have been installed since initial plant construction to control the reactor vessel head vent throttle valves. The throttling valves are used to vent non-condensable gases/steam from the reactor coolant system (RCS). The reactor vessel head vent throttle valves are designed to mitigate possible conditions of inadequate core cooling, inadequate natural circulation, or inability to depressurize the accumulation of non-condensable gases in the RCS. Additionally, the system may also be used as a backup RCS letdown flowpath to prevent filling the pressurizer solid after the reactor is tripped with normal and excess letdown unavailable. The Unit 1 controllers are located within panel 1-M-4 in the main control room (MCR), and the Unit 2 controllers are located within panel 2-M-4 in the MCR.

NEI 96-07, Appendix D, Supplemental Guidance for Application of 10 CFR 50.59 to Digital Modifications, Section 4.3.1 provides guidance, for the digital aspects of the proposed change, to apply use of a qualitative assessment approach to determine the answers to the Evaluation questions.

The change activity completed a qualitative assessment for the intended application of Yokogawa model YS1700 single loop controllers for the reactor head vent control valves in accordance with the guidance of NEI 96-07 and concluded as follows:

Based on the aggregate results, the qualitative assessment determined that there is sufficiently low likelihood of failure of the new digital equipment; therefore, the new Yokogawa YS1700 HIC is of equal or greater reliability than the equipment they replace. Therefore, it is concluded that any changes in accident frequency or equipment malfunction frequency meets the sufficiently low threshold described in NEI 96-07 Appendix D.

E1-4 These controllers are SR, Class 1E, and Seismic Category I devices.

The reactor vessel head vent throttle valves are used in conjunction with the reactor vessel head vent isolation valves to provide a vent pathway from the reactor vessel head to the pressurizer relief tank (PRT). One isolation valve and one throttle valve must be open at a minimum to allow flow to the PRT. When venting is desired, at least one of the isolation valves shall be opened by the operator via hand switches in the MCR. This is followed by utilizing HIC hand stations in the MCR to open at least one of the throttle valves. The existing Veritrak 75HC-3000 controllers have thumbwheels on the front right side that allow the operator to adjust the output signal to the valve (0-100%). The controller demand output on the existing controller is displayed on the front left side and the indication of the actual detected valve position on the front right side.

The Veritrak 75HC-3000 controllers are obsolete and replacement parts have become increasingly difficult to obtain.

The reactor vessel head vent controllers are needed to meet the Technical Requirement Manual (TRM) TR 8.4.3 requiring two RCSHV paths to be functional during Modes 1, 2 and 3.

The increasing likelihood of failures for the Veritrak 75HC-3000 controllers increases the risk that a TRM TR 7.5.3 condition will be entered and can't be restored in required TR 8.4.3 Restoration Time. This modification replaces these Veritrak 75HC-3000 reactor vessel head vent controllers with Yokogawa YS1700 controllers. The source of power for the controllers is not being changed with this modification.

The Yokogawa YS1700 controllers are configured to have the same throttle function and have the same functionality as the existing controllers with the only differences being the change from analog to digital (under normal circumstances) and a change in operation from the use of a thumbwheel to the use of directional buttons on the front of the controller. The human factors engineering evaluation completed within this proposed With respect to the scope of the 10 CFR 50.59 screened activities, the 10 CFR 50.59 Evaluation questions have been evaluated, all with an answer of NO. The screened in activities do not result in more than a minimal frequency of an accident previously evaluated in the UFSAR. The screened in activities do not result in a more than minimal impact to the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

The screened in activities do not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR as the restricting orifices are not altered by this change. The screened in activities do not result in a more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR as the restricting orifices are not altered by this change. The screened in activities do not create a possibility for an accident of a different type than any previously evaluated in the UFSAR. The screened in activities do not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered. Finally, the screened in activities do not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

Therefore, this activity may be implemented per plant procedures and does not require a license amendment request.

E1-5 activity documents that Yokogawa YS1700 has redundancy and furthermore can complete manual control functions even if both processors were to fail via analog circuits. The new controllers have the option of using a manual thumbwheel located behind the front panel of the controller if necessary and can be used in the event the processors were to fail. The replacement controllers utilize the same scale ranges and engineering units. A 4.0 milliampere (mA) signal will continue to represent a fully closed valve and a 20.0 mA signal will continue to represent a fully open valve.

The following aspect of the proposed modification was considered potentially adverse. The digital upgrade did not meet the criteria of a relatively simple digital upgrade and furthermore the failure of digital modification could result in a potential loss of display function. Loss of the display function would be adverse to the design function.

In accordance with the definitions of a design function from Nuclear Energy Institute (NEI) 96-07 Rev. 1, Guideline for 10 CFR 50.59 Implementation, the associated design function of the Unit 1 and Unit 2 reactor vessel head vent throttle valve controllers is to receive and send control signals so that the reactor vessel head vent throttle valves can be remotely operated from the MCR to both vent non-condensable gases or steam from the reactor vessel head and to function as a backup RCS letdown flowpath to prevent filling the pressurizer solid after the reactor is tripped with normal and excess letdown unavailable. Unintended mis-operation of the reactor head vent throttle valves could potentially result in an accident similar to a Loss of Reactor Coolant from Small Ruptured Pipes or Cracks in Large Pipes accident as described in UFSAR Section 15.3.1.

E1-6 Accept-As-is DESCRIPTION SAFETY ANALYSIS 1952024 and 1955859 The proposed change is an Accept As-is evaluation for Condition Reports (CRs) 1952024 and 1955859. The condition being evaluated is the possibility of foreign material being left in the Unit 2 Reactor Vessel at the conclusion of Unit 2 Refueling Outage 26 (U2R26).

CR 1952024 identifies debris observed in location K-08 during the control rod drive mechanism (CRDM) latch assembly video inspection. There was one small fragment located on the 180° stationary gripper (SG) latch arm tip, and two additional smaller fragments located at the top of the CRDM latch assembly (LA) guide tube. There was an attempt to retrieve the material during the outage via vacuum, but due to the small size of the debris it is assumed to still be within the CRDM for the purpose of this review. These pieces of foreign material cannot be classified as to its material composition, size, or thickness from the video inspection; however, Westinghouse provided an evaluation of the debris items. The material is conservatively assumed to be 304, 316, or 410 stainless steel, or Cobalt Alloy material, or Westinghouse PD-Spec 13206AA.

These materials were provided in the Westinghouse Loose Parts Assessment (LPA) which dispositioned these three items (Items #1, #2, and #3). The CR notes that the foreign material would not be expected to stop the CRDM LA from operating, nor performing its intended function of dropping the Drive Line when the reactor trip breakers are opened.

CR 1955859 documents the results of the U2R26 pre-reload lower core plate inspection and below core plate inspection. A camera was lowered down a flow hole in core location H-08 for the below core plate inspection. A piece of potential dark debris on the bottom support forging was identified. The articulated removal tool (ART) system was deployed to attempt debris retrieval below the core plate. After vacuuming debris at the bottom of the core barrel first, the bottom support forging was then thoroughly searched from multiple angles, but the potential dark debris could not be found again. Therefore, this The three metallic items and the unknown non-metallic item are bounded by multiple debris analyses performed in past evaluations. As such, this change is not considered adverse for the majority of design functions; however, due to the potential of the metallic debris items seen in the CRDM at location K08 to impede rod cluster control assembly (RCCA) insertion, this activity is considered adverse to CRDM operation and RCCA insertion.

These four items are conservatively assumed to be present in the RCS during all modes of operation following U2R26 outage. These items may circulate through the RCS and its connected systems during all modes of operation and for all design basis accidents. Therefore, all design basis accidents (DBAs) are involved in the acceptance of this material in the RCS. However, as discussed in the next section, there are no credible failures of structures, systems, or components (SSCs) that could be caused by this material so in this respect, there are no design basis accidents that are affected by this material being in the RCS.

There are no credible failures of SSCs that could be caused by these four debris items primarily because the items have insufficient mass, volume, and strength to:

Damage pressure boundary components or seal and/or gaskets between pressure boundary components.

Damage internal components in the RCS or connected systems or prevent their movement, most significantly, the fuel clad, pump impellers, valves, and the control rod drive mechanisms.

Degrade flow through the steam generator or any heat exchanger to the extent that performance of these components would be less than credited in the safety analysis.

Block instrument lines or interfere with the proper operation of any RCS or connected system instrumentation.

Block or degrade flow in the fuel assemblies to the extent that their performance or thermal hydraulic characteristics would be less than credited in the safety analysis.

Affect RCS chemistry.

Challenge CRDM operation or RCCA insertion capability.

E1-7 item (Item #4) is conservatively assumed to be unretrieved debris on the top of the Bottom Support Forging. The item is dark and non-metallic and of unknown composition.

Item #1: metallic item, dimensions less than 1 x 0.03x 0.5 [

equals inch]

Item #2 & #3: metallic items, dimensions less than 0.067 x 0.33 x 0.33 Item #4: non-metallic items, dimensions less than 0.41 x 0.41 x 0.05 The proposed activity is an Accept As-is review for leaving this potential foreign material or debris items in the Unit 2 Reactor Vessel.

Further, past operation experience at SQN has shown that debris of this type in the RCS does not cause equipment problems, malfunctions, or degradation.

This means the acceptance of this material in the RCS will not prevent the RCS or any system that connects to the RCS from performing its design functions in mitigating a design basis accident or in maintaining safe shutdown.

The evaluation concludes that the foreign material potentially left in the Unit 2 Reactor Vessel at the restart from the U2R26 outage will not adversely affect nuclear safety.

Therefore, this activity may be implemented per plant procedures and does not require a license amendment request.

Technical Requirement Manual DESCRIPTION SAFETY ANALYSIS 23-12 Technical Requirement (TR) 8.1.3, Rod Position Indication System - Shutdown, currently requires group demand position indication to be FUNCTIONAL and capable of determining within +/- 2 steps the demand position of each shutdown or control rod which is NOT fully inserted. This TR is applicable in Mode 3 (Hot Standby), 4 (Hot Shutdown), and 5 (Cold Shutdown) with the reactor trip breakers in the closed position.

If any required group demand position indicator is NOT functional, the prescribed Contingency Measure (CM) currently requires opening the reactor trip breakers immediately.

The proposed change to TR 8.1.3 consists of the following parts:

1. The acceptable (allowed) tolerance for group demand position indication is relaxed to +/- 12 steps. This change is consistent with the Technical Specification requirements for rod alignment and rod position indicator accuracy (applicable in Mode 1 or 2).

2. The CM for Condition A (less than required group demand position indicators FUNCTIONAL) is modified to allow 15 This 50.59 Evaluation concludes that the proposed TRM change does not impact the frequency or likelihood of any assumed accident or malfunction of an SSC because this activity only involves rod position indication requirements during shutdown conditions. Similarly, this change does not create the possibility of an accident of a different type or a malfunction of an SSC important to safety with a different result than previously evaluated in the UFSAR, as this does not change the function of the rod position indication nor physically modify the indication. This change does not increase the consequences of any assumed accident or malfunction of an SSC because the RCCA misalignment accident has no consequences when initiated during shutdown conditions. This activity does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered and does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

Therefore, this activity may be implemented per plant procedures and does not require a license amendment request.

E1-8 Technical Requirement Manual DESCRIPTION SAFETY ANALYSIS minutes to restore the required group demand position indicator(s) to FUNCTIONAL status.

3. New Condition B is added which applies if the CM and associated Restoration Time of Condition A is NOT met.

The CM for Condition B requires immediately initiating action to fully insert all rods OR immediately opening the reactor trip breakers. The option to fully insert all rods (in lieu of opening the trip breakers) is consistent with the existing wording of TR 8.1.3, which only requires demand position for each shutdown or control rod which is NOT fully inserted. Therefore, the alternate CM (fully inserting all rods) restores compliance with the TR.

4. The prescribed frequency for Technical Requirements Verification (TRV) 8.1.3.1 (which verifies functionality of group demand position indication by rod movement) is modified to allow completing this TRV within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after closing the reactor trip breakers if the TRV has NOT been completed within the previous 31 days. The existing 31 day frequency applies thereafter. This change recognizes that the reactor trip breakers must be closed to allow moving rods to verify functionality of the group demand position indicators and provides a reasonable time to complete this TRV after closing the trip breakers.

The proposed TRM changes screen in for a full evaluation due to adversely impacting the ability to monitor the status of rod positions and the resulting impact on shutdown core reactivity.

The group demand position indicators are NOT assumed by the FSAR accident analyses to be FUNCTIONAL in Modes 3-5, and are NOT relied upon to mitigate the consequences of a DBA or transient during shutdown conditions.

E1-9 DOCUMENT NUMBER/72.48 EVALUATION TRACKING NUMBER DESCRIPTION SAFETY ANALYSIS None

E2-1 ENCLOSURE 2 SEQUOYAH NUCLEAR PLANT COMMITMENT CHANGE REPORT Commitment Evaluation No./

Commitment Tracking No.

Source Document Summary of Original Commitment Summary of Commitment Changes Basis/Justification for Changes NCO930202003 (Unit 1)

NCO930202004 (Unit 2)

NRC letter to TVA dated March 15, 1994, Issuance of Amendments (TAC Nos.

M85308 and M85309)

(TS 92-16)

The Unit 1, in-service inspection (lSI) program will be revised by August 1, 1994, to incorporate the following requirements:

1) The ultrasonic technique for future augmented examinations of the reactor pressure vessel nozzles will be at least as sensitive as that used to conduct the examination during the Unit 1 Cycle 6 refueling outage;

2) The Unit 1 Cycle 6 (U1C6) examination will serve as the baseline for future examinations;

3) The augmented examination will be performed near the end of the second 10-year lSI interval for Unit 1; 4) Size all the detected flaws, regardless of the percent distance amplitude curve (DAC); 5) The results of the examinations will be submitted to NRC; 6) The augmented examinations will not be removed from the lSI program without notifying NRC.

The SQN ISI program will follow the examination requirements in ASME Section XI, Categories B-D and B-F, (as mandated by 10 CFR 50.55a) or approved alternatives per Regulatory Guide 1.147.

The license amendment for Unit 1 (Amendment No 177) and Unit 2 (Amendment No. 168) describes these nozzle exams as necessary to monitor the growth of cold cracks (Unit 1) and reheat cracks (Unit 2) under the nozzle cladding that was observed during construction exams.

Exam results from U1C6 (1993) and Unit 1 Cycle 14 refueling outage (U1C14)

(2006) showed no appreciable change in the sizing of the observed cracks and all indications were acceptable. The ASME Section XI Code, as mandated by 10 CFR 50.55a, has relaxed the exam requirements to the current state of only requiring a visual exam of nozzles inside radius sections per Code Case N-648-1.

Exam results from U2C6 (1994) and Unit 2 Cycle 13 refueling outage (U2C13)

(2005) showed no appreciable change in the sizing of the observed cracks and all indications were acceptable. The ASME Section XI Code, as mandated by 10 CFR 50.55a, has relaxed the exam

E2-2 Commitment Evaluation No./

Commitment Tracking No.

Source Document Summary of Original Commitment Summary of Commitment Changes Basis/Justification for Changes The Unit 2, in-service inspection (lSI) program will be revised by August 1, 1994, to incorporate the following requirements:

1) The volumetric examinations of the reactor pressure vessel nozzles will be performed over the same cladded nozzle areas required by American Society of Mechanical Engineers code;

2) The ultrasonic technique for the Unit 2 Cycle 6 (U2C6) refueling outage examinations and future examinations will be at least as sensitive as that used to conduct the examination during the Unit 1 Cycle 6 refueling outage; 3) The examinations to be performed during the U2C6 refueling outage will serve as the baseline for future examinations; 4) All of the detected flaws, will be sized regardless of the percent distance amplitude curve (DAC);

5) The results of the examinations will be submitted to NRC; 6) The above commitments will not be removed from the Unit 2 lSI program without notifying NRC.

requirements to the current state of only requiring a visual exam of nozzles inside radius sections per Code Case N-648-1.