Text

Design Basis Assurance Inspection (Teams) Inspection Report 05000327/2025010 and 05000328/2025010
	ML25084A085
Person / Time
Site:	Sequoyah
Issue date:	03/26/2025
From:	Renee Taylor; NRC/RGN-II/DORS/EB1
To:	Erb D; Tennessee Valley Authority
References
	IR 2025010
	Download: ML25084A085 (1)
	v • d • e

SUBJECT:

SEQUOYAH - DESIGN BASIS ASSURANCE INSPECTION (TEAMS)

INSPECTION REPORT 05000327/2025010 AND 05000328/2025010

Dear Delson Erb:

On February 13, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Sequoyah and discussed the results of this inspection with Tom Marshall and other members of your staff. The results of this inspection are documented in the enclosed report.

One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements and was determined to be Severity Level IV. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.

If you contest the violation or the significance or severity of the violation documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement; and the NRC Resident Inspector at Sequoyah.

If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; and the NRC Resident Inspector at Sequoyah.

March 26, 2025 Enclosure This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely, Ryan C. Taylor, Chief Engineering Branch 1 Division of Operating Reactor Safety Docket Nos. 05000327 and 05000328 License Nos. DPR-77 and DPR-79

Enclosure:

As stated

Inspection Report

Docket Numbers:

05000327 and 05000328

License Numbers:

DPR-77 and DPR-79

Report Numbers:

05000327/2025010 and 05000328/2025010

Enterprise Identifier:

I-2025-010-0027

Licensee:

Tennessee Valley Authority

Facility:

Sequoyah

Location:

Soddy-Daisy, TN

Inspection Dates:

January 27, 2025 to February 13, 2025

Inspectors:

P. Braxton, Reactor Inspector

J. Copeland, Reactor Inspector

L. Day, Reactor Inspector

C. Franklin, Reactor Inspector

J. Lizardi-Barreto, Reactor Inspector

A. Ruh, Senior Reactor Inspector

T. Su, Senior Reactor Inspector

Approved By:

Ryan C. Taylor, Chief

Engineering Branch 1

Division of Operating Reactor Safety

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a design basis assurance inspection (teams) inspection at Sequoyah, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors.

Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.

List of Findings and Violations

Failure to Obtain License Amendment Following Change to Reactor Vessel Head Vent Valve Usage Cornerstone Significance/Severity Cross-Cutting Aspect Report Section Mitigating Systems Green Severity Level IV NCV 05000328,05000327/2025010-01 Open/Closed

[H.13] -

Consistent Process 71111.21M The inspectors identified a Green finding and associated Severity Level IV NCV of 10 CFR 50.59(c)(1)(i) when the licensee failed to obtain a license amendment to change the technical specifications (TS) prior to utilizing reactor head vent valves as the means to prevent an accident from challenging the integrity of the reactor coolant system barrier.

Additional Tracking Items

None.

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.

REACTOR SAFETY

===71111.21M - Comprehensive Engineering Team Inspection The inspectors evaluated the following components and listed applicable attributes, permanent modifications, and operating experience:

Structures, Systems, and Components (SSCs) (IP section 03.01)===

Main article: IP 71111.21

For each component sample, the inspectors reviewed the licensing and design bases including:

(1) the Updated Final Safety Analysis Report (UFSAR);

(2) the Technical Specifications (TS); and

(3) the Technical Requirements Manual (TRM). The inspectors reviewed a sample of operating procedures (including normal, abnormal, and emergency procedures) and overall system/component health (including condition reports and operability evaluations, if any). The inspectors performed visual inspections of the accessible components to identify potential hazards and/or signs of degradation. Additional component specific design attributes reviewed by the inspectors are listed below.

(1) Essential raw cooling water pumps:

1. Maintenance effectiveness

2. Modifications

3. Translation of vendor specifications

4. Protection against seismic events

5. Flow capacity, flow balance, minimum flow, runout flow, required

submergence (NPSH, vortexing)

6. Test/inspection procedures, acceptance criteria, and recent results: Pump

comprehensive and quarterly inservice tests, emergency water make-up

7. Pump motor voltage drop, degraded voltage, brake horsepower, protective

devices, minimum voltage, control logic

(2) Essential raw cooling water piping:

1. Maintenance effectiveness

2. Modifications

3. Translation of vendor specifications

4. Protection against seismic events

5. Flow capacity, flow balance, hydraulic modeling, minimum flow

6. Test/inspection procedures, acceptance criteria, and recent results:

performance testing, leak-rate testing, aging management procedures/inspections

(3) Essential raw cooling water screenwash pumps:

1. Compliance with UFSAR, TS and TS Bases

2. Design basis documents and calculations

3. Evaluation of applicable operating experience relevant to sample, including

system health reports, maintenance records, and surveillance testing results

4. Visual inspection during walkdown focusing on material condition, operating

environment and component configuration

(4) Essential raw cooling water pumping station and access cells:

1. Maintenance effectiveness

2. Design review of design requirements

3. Operating condition design and licensing bases requirements

4. Operating procedures review

(5) Refueling water storage tank suction motor operated valve 1-LCV-62-135:

1. Operating procedures (including normal, abnormal, and emergency

procedures)

2. Modifications

3. Environmental qualification

4. Weak link analysis, closure and opening time, maximum allowed leakage

5. Test and inspection procedures, acceptance criteria, and recent results:

Leakage, In-service testing, thermal overload bypass testing, preventive maintenance

6. Motor voltage drop, control logic, required minimum voltage, degraded voltage

effects, brake horsepower, thermal overload protection, cable ampacity

(6) 125Vdc Battery II:

1. Operating environment

2. Material condition and installed configuration

3. System health reports

4. Surveillance test results

5. Calculations verifying system design requirements

(7) Shutdown board room chillers:

1. Licensing and design basis documents and calculations

2. Evaluation of applicable operating experience relevant to sample, including

system health reports, maintenance records, and test results

3. Visual inspection during walkdown focusing on material condition, operating

environment and component configuration

(8) 1A containment spray heat exchanger:

1. Engineering evaluations and visual inspections of structural integrity following

exposure to excessive differential pressure

2. Impact of tube plugging on expected thermal performance

3. Magnitude and impact of potential bypass flow conditions on thermal

performance

4. Wet layup and routine verification methods

Modifications (IP section 03.02) (4 Samples)

(1) SQN-20-1620, Rev. 1, Reactor vessel head vent controller replacement

(2) SQN-22-046, Rev. 0, Time critical action for spurious emergency core cooling system safety injection

(3) SQN-20-1132, Rev. 5, Non-conforming flood seals

(4) SQN-23-063, Rev. 0, Auxiliary feedwater level control valve seismic restraint splice detail 10 CFR 50.59 Evaluations/Screening (IP section 03.03) (8 Samples)

(1) Evaluation for SQN-20-1620, Rev. 0, Reactor vessel head vent controller replacement

(2) Evaluation for SQN-22-046, Rev. 0, Time critical action for spurious emergency core cooling system safety injection

(3) Evaluation of SQN-20-1174, Rev. 0, Residual heat removal minimum flow valve circuit changes due to fire scenarios

(4) Evaluation for SQN-20-1239-A, Rev. 1, Fuel transition changes

(5) Screening for SQN-24-001, Rev. 0, Change of auxiliary feedwater level control valves to be normally open

(6) Screening for SQN-23-046, Rev. 0, Review of a digital timing relay as a qualified replacement

(7) Screening of SQN-20-1684, Rev. 0, Feedwater heater level control upgrade

(8) Screening for SQN-19-885, Rev. 1, 6.9kV ITE breaker replacement equivalency

Operating Experience Samples (IP section 03.04) (2 Samples)

(1) NRC Information Notice 2013-05: Battery Expected Life and Its Potential Impact on Surveillance Requirements

(2) NRC Information Notice 2006-17: Recent Operating Experience of Service Water Systems Due to External Conditions

INSPECTION RESULTS

Failure to Obtain License Amendment Following Change to Reactor Vessel Head Vent Valve Usage Cornerstone Significance/Severity Cross-Cutting Aspect Report Section Mitigating Systems Green Severity Level IV NCV 05000328,05000327/2025010-01

[H.13] -

Consistent Process 71111.21M Open/Closed The inspectors identified a Green finding and associated Severity Level IV NCV of 10 CFR 50.59(c)(1)(i) when the licensee failed to obtain a license amendment to change the technical specifications (TS) prior to utilizing reactor head vent valves as the means to prevent an accident from challenging the integrity of the reactor coolant system barrier.

Description:

UFSAR section 15.2.14 evaluates a spurious operation of the safety injection (SI) system at power and concluded that the event presented no hazard to the integrity of the reactor coolant system and that the event would not degrade into a more serious plant condition. This result was possible because the time required to fill the pressurizer was demonstrated to exceed the required operator action time to terminate safety injection flow and establish a letdown flowpath. This operator action was necessary because a spurious safety injection signal, a Condition II event, can become a Condition III event (Small Break LOCA), if the resulting safety injection flow fills the pressurizer and a pressurizer relief or safety valve opens, discharges water, and then fails to close. This event escalation could be precluded if operator actions to terminate safety injection flow and establish a letdown flowpath can be completed before the pressurizer becomes water solid. The licensee discovered in 2023 that although the E-0 (Reactor Trip or Safety Injection) procedure would terminate SI within 15 minutes, the loss of letdown and continued flow from reactor coolant pump seal injection could still lead to overfilling the pressurizer if a letdown path was not established within a certain timeframe.

Design Change Package (SQN-22-046, Rev 0) established a new time-critical action (TCA) to establish a letdown source to preclude this event escalation based on pressurizer level indication. Following implementation of the change, UFSAR section 5.5.15 described that operation of the reactor vessel head vent valves was the credited letdown flow path for this event. This new TCA stated that a letdown flow path should be established within a conservative time of 5 minutes after pressurizer level is >80%. Periodic validation of this TCA using the simulator was most recently performed during the summer of 2024, with 5 separate crews successfully performing the timed events.

Although the modification appeared effective, the inspectors noted that the 10 CFR 50.59 screening concluded that a change to the TS was not necessary. Justification from the screening, in part, stated: the proposed activity consists of revisions to the TCAs involved with mitigating various single failures postulated to occur during a spurious SI at power event and the resulting change of the event incorporated revised inputs; however, all values defined by TS were unchanged. In summary, no TS are impacted by this change. Although the above assessment considered the existing TS, it did not also consider the possibility that an addition to the TS may be necessary to utilize the reactor head vent valves to mitigate an accident condition. Historically, in 2005, license amendment numbers 305 and 295 (ML052060033) relocated the reactor coolant system head vents from the former TS 3.4.11 into the technical requirements manual (TRM). This change was acceptable at the time because the RCS head vents were only used to assist in creating conditions conducive to natural circulation and were not components satisfying criterion 3 of 10 CFR 50.36(c)(2)(ii).

10 CFR 50.36(c)(2) includes the criteria for establishment of TS LCOs and defines limiting conditions for operation (LCO) as the lowest functional capability or performance levels of equipment required for safe operation of the facility. Criterion 3 in 50.36(c)(2)(ii)(C) required a TS LCO be established for a component that is part of the primary success path and which actuates to mitigate a design basis accident that presents a challenge to the integrity of a fission product barrier. Following SQN-22-046, the safety analyses were revised to use the reactor vessel head vent valves as the means to ensure a Spurious Operation of the Safety Injection System at Power accident did not escalate to a small break loss of coolant accident.

Inspectors noted that Sequoyah's procedure NPG-SPP-09.4 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments has historically lacked guidance to ensure screenings of plant changes would reliably recognize when activities require a change to the TS. In the most recent version, Attachment 1, section 1.5 gave guidance for when a plant change credited new components in a safety analysis. It stated that since 10 CFR 50.36(c)(2)(iii)does not require adding new [components] to the TSs for those licensees that had an operating license on or before August 18, 1995, that Sequoyah may not need to add a new TS for a component that met the criteria in 10 CFR 50.36. This exclusion granted in 50.36(c)(2)(iii) was to preclude backfitting the criteria of 50.36(c)(2)(ii) into licensees pre-existing TS and was not appropriate guidance for screening whether plant changes implemented after August 18, 1995 required a change to the TS.

Despite Sequoyah not obtaining a licensee amendment to add a TS LCO for the reactor vessel head vent valves, the inspectors confirmed that the licensee was monitoring the functionality of these valves per the TRM. The inspectors reviewed the Sequoyah plant logs and noted that the individual trains of equipment had been non-functional for no more than approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> since implementation of the change.

Corrective Actions: The licensee confirmed that the reactor vessel head vents were functional, available, and being administratively controlled per TRM 8.4.3 while evaluations were conducted to determine if more restrictive administrative controls were needed in the interim.

Corrective Action References: 1988736

Performance Assessment:

Performance Deficiency: The failure to obtain a license amendment to make a change to the TS as required by 10 CFR 50.59(c)(1)(i) prior to implementing modification SQN-22-046 was a performance deficiency. Specifically, criterion 3 in 50.36(c)(2)(ii)(C) required a TS LCO be established for a component that is part of the primary success path and which actuates to mitigate a design basis accident that presents a challenge to the integrity of a fission product barrier.

Screening: The inspectors determined the performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, without a TS LCO being established, the plant could be unknowingly operated without the minimum equipment needed to ensure the safety analyses remained bounding.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Exhibit 2 "Mitigating Systems Screening Questions," inspectors determined the issue was Green because the inspectors confirmed that the regulatory process error had not led to a physical degraded condition where the necessary equipment was unavailable for any significant periods of time (i.e. greater than 3 days).

Cross-Cutting Aspect: H.13 - Consistent Process: Individuals use a consistent, systematic approach to make decisions. Risk insights are incorporated as appropriate. In this case, the licensees historical guidance in NPG-SPP-09.4 lacked a consistent, systematic approach for ensuring screenings of plant changes would reliably recognize when activities required a change to the TS. More recent revisions had enhanced guidance that might have prevented the present issue, however, the present guidance included conflicting statements that suggested Sequoyah may be excluded from needing to add new TSs for SSCs meeting the criteria of 10 CFR 50.36.

Enforcement:

The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance.

Severity: Based on the examples provided in section 6.1 of the Enforcement Policy, dated January 14, 2022, "Reactor Operations," the issue was determined to be a SL IV violation.

Specifically, example 6.1.d.2 states that a SL IV violation involves violations of 10 CFR 50.59 resulting in conditions evaluated as having a very low safety significance (i.e. green) by the significance determination process.

Violation: 10 CFR 50.59(c)(1)(i) requires, in part, that the licensee may make changes without obtaining a license amendment only if a change to the TS is not required. 10 CFR 50.36(c) established what items are necessary to include in TS, and 50.36(c)(2)(ii)(C)required a TS LCO be established for a component that is part of the primary success path and which actuates to mitigate a design basis accident that presents a challenge to the integrity of a fission product barrier. Contrary to the above, the station failed to obtain a license amendment to add a TS LCO after making a change to utilize reactor head vent valves as the means to prevent an accident from challenging the integrity of the reactor coolant system barrier. Specifically, following modification SQN-22-046, the safety analyses in UFSAR sections 5.5.15 and 15.2.14 were revised to use the reactor vessel head vent valves as the means to ensure a Spurious Operation of the Safety Injection System at Power accident did not escalate to a small break loss of coolant accident.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

EXIT MEETINGS AND DEBRIEFS

The inspectors verified no proprietary information was retained or documented in this report.

On February 13, 2025, the inspectors presented the design basis assurance inspection (teams) inspection results to Tom Marshall and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

00D53EPMRJP061091

Generic Letter 89-10 MOV Population at

Sequoyah Units 1 & 2

1-LCV-62-135

Documentation of Design Basis Review, Required

Torque Calc and Valve & Actuator Capability

Assessment for 1-LCV-62-135

31D53EPMGDF01030687

HVAC Cooling Load Calculation: Aux Bldg Board

Room and Shutdown Board Room

2D530HCGTBG082181

Containment Spray Heat Exchanger UA Value and

Tube Plugging Limits

CEBCQS364

Shutdown Board Room Chillers

EDQ0009992017000385FHA

Sequoyah Nuclear Plant - Fire Hazard Analysis

Calculation - FSSD Compliance Mitigation

Strategy

MDQ0000622022000214

Mov Differential Pressure Calculation - Chemical

Volume Control (CVCS) System MOVS

MDQ00006720000095

ERCW Flow Balanced Hydraulic Model

MDQ00006720020109

ERCW System Sensitivity Review for 87F, ESF, &

HVAC Equipment

MDQ0010722024000000

Containment Spray Heat Exchanger 1A UA Value

and Tube Plugging Limits

MDQ0067970004

ERCW Protoflo Hydraulic Model

SQN-APS-003

480VAC APS Class 1E Load Coordination Study

27

SQN-APS-003

480VAC APS Class 1E Load Coordination Study

27

SQN-CPS-013

Analysis Of Control Circuit Fuses For 120V AND

25V DC Systems

SQN-CPS-051

Circuit Protection Device Evaluation

SQN-CPS-058

Vital Control Power System Loading Channel II

and Continuous Loading Evaluation of Protective

Devices in the 120V

AC Vital Instrument Power Boards

135

SQN-EEB-MS-TI06-0008

Degraded Voltage Analysis

71111.21M

Calculations

SQN-VD-VDC-1

25 VDC Vital Instrument Power System Design

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

Verification

SQNDES1003

ERCW Screen Wash System Hydraulic Analysis

and ERCW Pump NPSHA Analysis

005

SQNETAPAC

Auxiliary Power Systems

24

SQTP002

ASME Section XI & OM Inservice Pump &

Augmented Pump Identification for The Second &

Third Ten Year Interval

Corrective Action

Documents

274, 85524, 1578600,

1609392, 1611076,

1638548, 1668113,

1691745, 1697633,

1733830, 1780764,

1788094, 1814813,

22187, 1847095,

1902377, 1780764,

1788094, 1814813,

22187, 1847095,

0405860, 0474374,

0906111, 1463884,

1788503, 1795467,

1983796, 1986983,

1742723, 1595633,

1911625, 1911944,

1753973, 702753, 1990322,

1804719, 1832893,

1832905, 1835609,

1837794, 1849221,

1851824, 1869442,

1918056, 1918316,

1918396, 1923909,

24797, 1924847,

27185, 1927437,

27846, 1929779,

1930216, 1942927,

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

1984365, 1988736,

1950945, 1988736,

22292, 1856592,

1778610, 1957182,

1942717, 1930867, 1928234

1984401

50.59 Documents not loaded to ECM

1987812

0-PI-SXX-027-694.0 needs cancellation

1987815

PCR on 0-TI-CEM-000-712.0

1987827

PCR for 0-TI-CEM-000-712.0

1987828

Recording error in 0-PI-CEM-000-460.4 from

8/13/24

1988163

Red Duct Tape Splash Guard on the M-B ERCW

Pump

1988172

House Keeping Issues at ERCW Building

1988398

Perform Calibration on A-A SDBR Chiller Make-Up

Water Valve

1988407

Calibrate A-A SDBR Chiller Chilled Water

Pressure Indicator PI-313-300

1988453

Heat trace junction box needs adjusting

1988642

2A ERCW (2-67-852B) Quick Disconnect Leak

1988736

Potential violation of 10CFR50.36

1988744

Transit scaffold issue

1988754

Remove writing on Penetration

1988762

Lost record. WO 121337046 could not be located.

1990322

FCR-TVA-2022-1777 Incorrectly Implemented

1991383

ECP SQN-20-1620 50.59 Evaluation Compliance

with NEI 96-07

1991538

PCR on EA-201-2

1991630

Enhancement Opportunity for Ferrite Beads

Discussion

1991635

Mode restraint not tracked as required

1991639

PCR on 0-GO-1

Corrective Action

Documents

Resulting from

Inspection

1991776

NRC Identified Sample IQ Typographical Error

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

1991895

Flood Seals

1991966

NRC Observation, Potential for by-pass flow not

considered in 1A CS HX evaluation

1,2-45N703-2

Wiring Diagram 125 Vital Battery Board II-Single

Line Sheet 2

1,2-45N751-6

Wiring Diagrams 480 V Reactor MOV BD 1B1-B

Single Line Sheet 2

1,2-47W852-3

Mechanical Flow Diagram-Floor and Equipment

Drains

1-37W206-10

Mechanical ERCW Pumping Station Piping and

Equipment

1-37W206-2

Mechanical ERCW Pumping Station Piping &

Equipment

1-47W811-1

P&ID: Safety Injection System

1-47WHYD-ERCW

External Flood Hazard Barriers ERCW Pumping

Station

2-37W206-10

Mechanical ERCW Pumping Station Piping and

Equipment

2-37W206-2

Mechanical ERCW Pumping Station Piping &

Equipment

2-45B640-39

Contact Development of Selector Switches and

Pushbuttons

2-45N2645-9

Wiring Diagrams Unit Control Board-PNL 2-M-6

Connection Diagrams-Sheet 9

2-47W809-1

P&ID: Chemical & Volume Control System

2-47W845-1

P&ID: Essential Raw Cooling Water System

2-47WHYD-ERCW

External Flood Hazard Barriers ERCW Pumping

Station

47W479-15

Mechanical Drains & Embedded Piping

C-12706.10

Containment Spray Heat Exchanger 1A Bill of

Material

D-12706.01

Containment Spray Heat Exchanger 1A Assembly

Drawings

D-12706.02

Containment Spray Heat Exchanger 1A Shell &

Channel Partial Section

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

D-12706.04

Containment Spray Heat Exchanger 1A Tube

Bundle

DCN 23322 A

Replace the 1A-A Containment Spray Heat

Exchanger

AA

SQN-19-885

6.9kV ITE Breaker Replacement Equivalency

SQN-20-1132

Non-conforming Flood Seals

SQN-20-1239-02

SQN-2 WEC RFA-2 Fuel Transition

SQN-20-1620

Reactor Vessel Head Vent Controller

Replacement

SQN-22-046

Time Critical Action for Spurious ECCS SI

SQN-23-063

AFW LCV Seismic Restraint Alignment Splice

Detail

SQN-24-001

Change Normal Configuration of Aux Feedwater

LCVs to Open

Engineering

Changes

SQN-24-013

Containment Spray Heat Exchanger 1A Tube

Plugging Limit Increase

2300527.401

1RF26 Risk Assessment of Degraded 1A

Containment Spray HX

04/18/2024

DC-SQN-20-1684

CFR 50.59 Screening Review

EWR 124437716

Evaluation of damage to the 1A CS Hx for TVA

SQN by Joseph Oat/SI

04/17/2024

S10211020800

TRM Change 21-08

SQN-20-1174

App R Concern - RHR Min Flow Valves

SQN-20-1174-04

App R Concern - RHR Min Flow Valve 2-FCV-74-

(Stage 4)

SQN-20-1684

Design Change Package: U1 FWH Level Control

Upgrade

SQN-22-030

ECP to Regain Margin for ERCW Screen Wash

Pumps

SQN-23-046

Digital Evaluation for Ametek Supplied Digital

Relay

SQN-24-025

50.69 System 311/313 Control/ Service Building

HVAC Risk Basis Document UNID Update

Engineering

Evaluations

TM-2891

Heat Exchanger Damage Evaluation

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

CNL-20-014

Application to Modify the Sequoyah Nuclear Plant

Units 1 and 2 Technical Specification to Allow for

Transition to Westinghouse RFA-2 Fuel (SQN-TS-

20-09)

09/23/2020

IN 2006-17

Recent Operating Experience of Service Water

Systems Due to External Conditions

03/06/2007

PR034998EMC-TR16

Qualification TestReport for EMC Testing YS1700

Programmable Indicating Controller

Raw Water Treatment

Program, April 2024