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Triennial Fire Protection Inspection Report 05000327/2023011 and 05000328/2023011
	ML23286A005
Person / Time
Site:	Sequoyah
Issue date:	10/16/2023
From:	Daniel Bacon; NRC/RGN-II/DRS/EB2
To:	Jim Barstow; Tennessee Valley Authority
References
	IR 2023011
	Download: ML23286A005 (1)
	v • d • e

SUBJECT:

SEQUOYAH NUCLEAR PLANT - TRIENNIAL FIRE PROTECTION INSPECTION REPORT 05000327/2023011 AND 05000328/2023011

Dear James Barstow:

On October 11, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Sequoyah Nuclear Plant and discussed the results of this inspection with Meshelle Augustin 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. 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 Nuclear Plant.

October 16, 2023 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, Daniel M. Bacon, Chief Engineering Branch 2 Division of 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:

05000327/2023011 and 05000328/2023011

Enterprise Identifier:

I-2023-011-0030

Licensee:

Tennessee Valley Authority

Facility:

Sequoyah Nuclear Plant

Location:

Soddy-Daisy, TN

Inspection Dates:

August 14, 2023, to September 01, 2023

Inspectors:

P. Braaten, Senior Reactor Inspector

E. Coffman, Reactor Inspector

J. Montgomery, Senior Reactor Inspector

D. Strickland, Reactor Inspector

Approved By:

Daniel M. Bacon, Chief

Engineering Branch 2

Division of Reactor Safety

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a triennial fire protection inspection at Sequoyah Nuclear Plant, 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

Reliance on OMAs In Lieu Of Physical Protection Without Approval or Evaluation Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000327,05000328/2023011-01 Open/Closed None (NPP)71111.21N.

Inspectors identified a Green finding and associated non-cited violation (NCV) of Sequoyah Nuclear Plant (SQN) Operating License Conditions (OLCs) 2.C(16) and 2.C(13), for Units 1 and 2 respectively, for the licensee's reliance on operator manual actions in lieu of protecting cables and equipment to ensure that one train of equipment required for SSD was free of fire damage.

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

71111.21N.05 - Fire Protection Team Inspection (FPTI) Structures, Systems, and Components (SSCs) Credited for Fire Prevention, Detection, Suppression, or Post-Fire Safe Shutdown Review (IP Section 03.01)

Main article: IP 71111.21

The inspectors verified that components and/or systems will function as required to support the credited functions stated for each sample. Additional inspection considerations are located in the fire hazards analysis (FHA) or safe shutdown analysis (SSA).

(1) Automatic preaction sprinkler system in FAA-091 (Auxiliary Control Instrument Room)

(2) Automatic preaction sprinkler system in FAA-094 (480V Board Room 1A)

(3) Unit 1 Chemical Volume & Control System

(4) Unit 1 6.9kV Electrical Distribution System

Fire Protection Program Administrative Controls (IP Section 03.02) (1 Sample)

The inspectors verified that the selected control or process is implemented in accordance with the licensees current licensing basis. If applicable, ensure that the licensees FPP contains adequate procedures to implement the selected administrative control. Verify that the selected administrative control meets the requirements of all committed industry standards.

(1) Fire Brigade Training Program

Fire Protection Program Changes/Modifications (IP Section 03.03) (2 Samples)

The inspectors verified the following:

a. Changes to the approved FPP do not constitute an adverse effect on the ability to safely shutdown.

b. The adequacy of the design modification, if applicable.

c. Assumptions and performance capability stated in the SSA have not been degraded through changes or modifications.

d. The FPP documents, such as the Updated Final Safety Analysis Report, fire protection report, FHA, and SSA were updated consistent with the FPP or design change.

e. Post-fire SSD operating procedures, such as abnormal operating procedures, affected by the modification were updated.

(1) SQN-20-1174, Appendix R Concern - RHR Min Flow Valves

(2) RIS 2005-07 Analysis for WO 120862019 MM/PI/0-PI-FPU-026-101.0 ATT 5 &

12/AUX 734 and 749 A-Train Sprinkler Inspection Sample

INSPECTION RESULTS

Reliance on OMAs In Lieu Of Physical Protection Without Approval or Evaluation Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000327,05000328/2023011-01 Open/Closed None (NPP)71111.21N.0 Inspectors identified a Green finding and associated non-cited violation (NCV) of Sequoyah Nuclear Plant (SQN) Operating License Conditions (OLCs) 2.C(16) and 2.C(13), for Units 1 and 2 respectively, for the licensee's reliance on operator manual actions in lieu of protecting cables and equipment to ensure that one train of equipment required for SSD was free of fire damage.

Description:

For fire safe shutdown, the Chemical Volume & Control System (CVCS) is credited to provide Reactor Coolant Pump seal integrity and Reactor Coolant System inventory control via the seal injection flow path.

During this inspection, the inspectors reviewed credited components in the CVCS for a sample of fire areas and noted, for the main charging flow control valve 1-FCV-062-093, the safe shutdown analysis stated, Resolution record to ensure MOA for local control of RCP seal injection flow can be credited. Inspectors also noted this justification was present for every fire area in the plant. The team questioned the licensing basis for crediting a manual operator action (MOA) (also known as operator manual action [OMA]) in lieu of protecting cables and equipment to ensure that one train of equipment required for safe shutdown (SSD) was free of fire damage. The licensee stated that calculation XDN00000020110001, Operator Manual Actions (OMAs) 10 CFR 50, Appendix R III.G.2 Compliance documented the feasibility, reliability, and defense-in-depth of the action to manually throttle seal injection.

However, after reviewing the plants licensing basis documents, inspectors determined that this action had never been reviewed and approved by the NRC, and that this condition had been noted by a previous inspection team.

In 2012, NRC inspectors issued the licensee a Green non-cited violation (NCV) for relying on OMAs in lieu of protecting cables and equipment to ensure that one train of equipment required for SSD was free of fire damage (Inspection Report 05000327, 328/2012-002). The OMAs that were the subject of the NCV were the following:

1. Disable affected emergency diesel generator (EDG) breakers and energize shutdown

boards from normal feeder within 30 minutes.

2. Place spare vital inverter in service within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3. Restore power to battery charger by transferring 480 Volt vital transfer switch to its

alternate switch position within 45 minutes.

4. Throttle reactor coolant pump seal injection flow to limit pressurizer level rise within 2

hours.

In the 2012 NCV, the inspectors stated "...the licensees methodology of allowing fire damage to occur (in lieu of protecting SSD cables and equipment) and relying on OMAs to achieve post-fire SSD would adversely affect SSD. Thus, adding the new OMAs to the sites fire protection program in 2002 required prior NRC approval because that change adversely affected the ability to achieve and maintain SSD in the event of a fire." The inspection report also stated that the evaluation performed in 2002 for the new OMAs was not adequate to support the conclusion that adding the OMAs did not adversely affect post-fire SSD because the evaluation only addressed OMA feasibility and did not address defense-in-depth.

During this inspection, the team reviewed PER 324757, which was written to address the issue noted above. The team determined that the licensees corrective actions were erroneously focused on re-performing an analysis to address defense-in-depth rather than addressing the fact that reliance on OMAs in lieu of protecting SSD cables and equipment is not allowed by the plants license unless reviewed and approved by the NRC, or the SSD cables and equipment have been demonstrated to not be required for SSD per guidance in NEI 00-01, Guidance for Post Fire Safe Shutdown Circuit Analysis; as partially endorsed by the NRC in RG 1.189, Fire Protection for Nuclear Power Plants.

NEI 00-01 and RG 1.189 describe two classifications of equipment in the plantequipment that is required to achieve and maintain hot-shutdown conditions and equipment that is important to safe shutdown. Equipment that is required to achieve and maintain hot-shutdown conditions must be physically protected to ensure that one train of equipment required for SSD is free of fire damage. Reliance on OMAs is not allowed for this category of equipment, unless reviewed and approved by the NRC. On the other hand, OMAs are an option for equipment that is important to safe shutdown. Licensees must have an analysis to demonstrate that an SSC is important to safe shutdown vice required to achieve and maintain hot-shutdown conditions.

Currently, the plants OMA calculation identifies actions (1), (2), and

(3) above as being important to safe shutdown. However, the licensee did not have an analysis to support the classification of the actions as being important to safe shutdown vice required for hot shutdown. RG 1.189 states, The information included in Appendix H to NEI 00-01, Revision 4, may be used in classifying components on the success path required for hot shutdown and those important to safe shutdown, when applied in conjunction with this RG. Further, while Appendix G to NEI 00-01, Revision 4, identifies suggested classifications for certain safe shutdown functions and SSCs, each licensee should perform their own plant-specific evaluations.

Additionally, the plants OMA calculation identifies action

(4) above as being required for safe shutdown. The plant erroneously believed that this action had been approved in a 1988 Safety Evaluation Report (SER). The team determined that this SER documented approval of other OMAs, and each of these OMAs were explicitly discussed. However, the SER did not explicitly discuss or document anything regarding throttling reactor coolant pump seal injection flow.

Corrective Action References: CR 1879011

Performance Assessment:

Performance Deficiency: Inspectors determined the licensee's failure to ensure that one of the redundant trains of equipment necessary to achieve and maintain hot shutdown conditions was free of fire damage was a performance deficiency. Specifically, the licensee relied on OMAs without NRC approval, or proper evaluation, in lieu of protecting cables and equipment.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the reliance on OMAs without prior NRC approval, or proper evaluation, adversely affected the ability to achieve and maintain safe shutdown in the event of a fire.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix F, Fire Protection and Post - Fire Safe Shutdown SDP. The issue was determined to impact Post-Fire Safe Shutdown and was not assigned a low degradation rating. For 3 of the 4 OMAs, the inspectors determined the issue was of very low safety significance (Green)because the finding did not adversely affect the ability to reach and maintain hot shutdown/hot standby conditions using the credited safe shutdown success path (Question 1.4.7-C). For the remaining OMA (throttle reactor coolant pump seal injection flow to limit pressurizer level rise within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ), the NRC utilized the licensee's fire PRA to assess the risk in accordance with section 1.5 of the SDP. The inspectors confirmed that the licensee has a NRC Regulatory Guide 1.200 compliant PRA, which has received a peer review, and all peer review comments have been addressed. The licensee performed a risk-based qualitative analysis because the OMA to throttle reactor coolant pump seal injection flow was not directly modeled in the fire PRA. The licensee's qualitative assessment determined the change in CDF would be zero because, per fire response operating procedures, operators would initiate plant cooldown within one hour of a fire event that necessitates plant shutdown. The reduction in RCS temperature and pressure associated with plant cooldown would alleviate the need to throttle seal injection flow to control RCS level. The SRA reviewed this evaluation and agreed with the licensees conclusion; therefore, inspectors determined the risk associated with the reactor coolant pump seal injection flow OMA was of very low safety significance (Green).

Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.

Enforcement:

Violation: Sequoyah OLCs 2.C(16) and 2.C(13) for Units 1 and 2 respectively, state in part, the licensee shall implement and maintain in effect all provisions of the approved fire protection program referenced in the SQN Final Safety Analysis Report (FSAR). Sequoyah FSAR Section 9.5.1, Fire Protection System, states that the fire protection system and fire protection features are described in the SQN Fire Protection Report (FPR), and the FPR should be referred to for a detailed description of the Fire Protection Plan. Section 2.0 of Part II of the SQN FPR states, in part, that the FPR describes the Fire Protection Plan developed for SQN to ensure compliance with the requirements of...10 CFR 50, Appendix R, Sections III.G, J, L, and O. 10 CFR 50, Appendix R, Section III.G.2, states, in part, that where cables or equipment, that could prevent operation or cause mal-operation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, one of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided:

(a) separation of cables and equipment by a fire barrier having a 3-hour rating,

(b) separation of cables and equipment by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards and with fire detectors and an automatic fire suppression system in the fire area, or

(c) enclosure of cables and equipment in a fire barrier having a 1-hour rating and with fire detectors and an automatic fire suppression system in the fire area.

Contrary to the above, the licensee failed to use one of the means described in Appendix R,Section III.G.2.a, b, or c to ensure that one of the redundant trains of equipment necessary to achieve and maintain hot shutdown conditions was protected from fire damage.

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 October 11, 2023, the inspectors presented the triennial fire protection inspection results to Meshelle Augustin and other members of the licensee staff.

On August 31, 2023, the inspectors presented the primary inspection results to Chris Reneau and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

v t e Inspection Report - Sequoyah - 2023011
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \| 25 \|
2023 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 10(0) 11(0) Security 401(0) 402(0) 403(0) 420(0) 440(0) 441(0) 442(0) Operator Exams 301(0) Emergency Planning Other

IR 05000327/2023011