IR 05000327/2021010

NRC Inspection Report 05000327/2021010 and 05000328/2021010

ML21112A344
Person / Time
Site:	Sequoyah
Issue date:	04/21/2021
From:	James Baptist NRC/RGN-II/DRS/EB1
To:	Jim Barstow Tennessee Valley Authority
References
IR 2021010
Download: ML21112A344 (14)

Text

April 21, 2021

SUBJECT:

SEQUOYAH NUCLEAR PLANT, UNITS 1 & 2 - NRC INSPECTION REPORT 05000327/2021010 AND 05000328/2021010

Dear Mr. Barstow:

On March 12, 2021, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Sequoyah Nuclear Plant, Units 1 & 2. On April 14, 2021, the NRC inspectors discussed the results of this inspection with Mr. John Taylor - Senior Manager SQN Design Engineering and other members of your staff. The results of this inspection are documented in the enclosed report.

No NRC-identified or self-revealing findings were identified during this inspection.

A licensee-identified violation which was determined to be of very low safety significance and Severity Level IV is documented in this report. 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, Units 1 & 2.

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,

/RA/

James B. Baptist, Chief Engineering Branch 1 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:

05000327/2021010 and 05000328/2021010

Licensee:

Tennessee Valley Authority

Facility:

Sequoyah Nuclear Plant, Units 1 & 2

Location:

Soddy Daisy, TN

Inspection Dates:

March 08, 2021 to March 12, 2021

Inspectors:

T. Fanelli, Sr. Construction Inspector

M. Greenleaf, Reactor Inspector

N. Morgan, Reactor Inspector

D. Terry-Ward, Construction Inspector

Approved By:

James B. Baptist, Chief

Engineering Branch 1

Division of Reactor Safety

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a NRC inspection at Sequoyah Nuclear Plant, Units 1 & 2, 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. A licensee-identified non-cited violation is documented in report section: 71111.17T.

List of Findings and Violations

No findings or violations of more than minor significance were identified.

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.

Starting on March 20, 2020, in response to the National Emergency declared by the President of the United States on the public health risks of the coronavirus (COVID-19), inspectors were directed to begin telework. In addition, regional baseline inspections were evaluated to determine if all or portion of the objectives and requirements stated in the IP could be performed remotely. If the inspections could be performed remotely, they were conducted per the applicable IP. In some cases, portions of an IP were completed remotely and on site. The inspections documented below met the objectives and requirements for completion of the IP.

REACTOR SAFETY

71111.17T - Evaluations of Changes, Tests, and Experiments Sample Selection (IP Section 02.01)

The inspectors reviewed the following evaluations, screenings, and/or applicability determinations for 10 CFR 50.59 from [enter dates].

(1)10 CFR 50.59 Evaluation - DCN SQN-19-308, Diesel Generator 2A-A VORD Removal, Rev. 2 (2)10 CFR 50.59 Evaluation - ECP 100098, NIS Time Delay Relay Project, Rev. 2 (3)10 CFR 50.59 Evaluation - DCN 100069, Upgrade the Turbine Driven Auxiliary Feedwater Speed Governor and Flow Controller, Rev. 3 (4)10 CFR 50.59 Evaluation - DCN D23216, 1,2-HS-062-0104C, -108C, -003-0118C, -

0128C, -068-0341AC, -0341DC, Admin. Change #04, Rev. A (5)10 CFR 50.59 Evaluation - D22501, Increase the Capability of 2-FCV-63-72 & -73 to Operate Under Differential Pressure, Rev. A (6)10 CFR 50.59 Evaluation - DCN D23085, Replace Breaker Handle, Breaker Operating Mechanism and Remove Kirk-Key Interlocks, Rev. A (7)10 CFR 50.59 Screen - D23623-01, Charging Flow Isolation Valve 1-FCV-062-0091-B, Rev. A (8)10 CFR 50.59 Screen - EC 23699-03, Change to Allow Flow Element Removal, Rev.

B (9)10 CFR 50.59 Screen - DCN 23885, Issue "Emergency and Abnormal Operating Procedure Setpoints" Calculation SQS20110, Revision 30, Design Output, Rev. 0 (10)10 CFR 50.59 Screen - DCN 22778, Auxiliary Feedwater Min-Flow Protection, Rev. A (11)10 CFR 50.59 Screen - DCN 21892, Install CCUS 2-XC-43-55, -58, -61, and -64 in the Circuit Before Solenoid Valves 2-FSV-43-55, -58, -61, and -64 to Regulate Current And Reduce Heat Rise, Rev. A (12)10 CFR 50.59 Screen - ECP 10068-01, Motor Operated Potentiometers (MOP)

Replacement, Rev. 2 (13)10 CFR 50.59 Screen - ECP SQN-19-785, ECP That Will Eliminate the Need for the Current SBO Diesel Generators by Crediting the FLEX Program, Rev. 1 (14)10 CFR 50.59 Screen - DCN SQN-19-856, Replace Obsolete EQ Transmitter 2-FT-3-163-B, Rev. 1 (15)10 CFR 50.59 Screen - DCN 23813, Replace Steam Generator Blowdown Containment Isolation Sample Valves 1- & 2-FSV-043-0055-B, -0058-A, -0061-B, -

0064-A, Rev. 5 (16)10 CFR 50.59 Screen - DEC SQN-19-699, EQ Binder Revisions for SG Level Transmitter 1-LT-003-0097-G, Rev. 0 (17)10 CFR 50.59 Screen - EDC E21192, WO 09-779231-000, Replace solenoid (per Engineering request), Per MI-10.38 and EDC E21192 with appropriate replacement, acceptance by OPS 6/27/19, Rev. A (18)10 CFR 50.59 Screen - DCN 22346, Replace 480V Board Room Air Conditioners &

Associated Equipment, Rev A (19)10 CFR 50.59 Screen - DEC SQN-18-226, Limit Switch Variation for 2-FCV-67-91, Rev. 1 (20)10 CFR 50.59 Screen - DEC 100075, B Train Shutdown Board Cleaning Maintenance Crosstie, Rev.

INSPECTION RESULTS

Licensee-Identified Non-Cited Violation 71111.17T This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Violation: The licensee performed design change DCN SQN-19-308 which removed the Voltage Overshoot Reduction Device (VORD) from the 2A-A Emergency Diesel Generator (EDG) after the VORD was damaged in 2015. The function of the VORD is to reduce the overvoltage developed by the EDG's voltage regulation system during sequence loading during design basis accidents. The DCN incorrectly determined that the VORD removal was acceptable with respect to their licensing and design bases acceptance criteria.

The station correctly determined that the licensee's commitment to Regulatory Guide (RG)1.9, "Selection, Design, and Qualification of Diesel-Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," was a commitment to a Revision 0 of RG 1.9 with the exception of the portion of the RG related to load sequencing. The licensee had been allowed by the NRC to deviate from Rev. 0's 40% load sequence interval acceptance criteria to Rev. 1's acceptance criteria of 60% (as described in NUREG-1232). The licensee determined that they had incorrectly determined that they were allowed to use a provision of RG 1.9 Rev. 1 to exceed the 60% acceptance criteria. To support the modification, the licensee also determined that they had incorrectly altered Section 8.3.1.2.1 of their Updated Final Safety Analysis Report (UFSAR) stating that: "A greater percentage of the time interval may be used if it can be justified by analysis."

10 CFR 50.59(c)(2)(ii) required, in part, that a licensee to obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change if the change would result in a more than minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the final safety analysis report (as updated).

Contrary to the above, since the completion of DCN SQN-19-308, Diesel Generator 2A-A VORD Removal, in June of 2019, the licensee failed to obtain a license amendment prior to implementing the DCN which allowed the emergency diesel generator to exceed its licensing basis acceptance criteria to which they were committed, thereby resulting in a more than minimal increase in the likelihood of occurrence of a malfunction of the emergency diesel generator 2A-A. Specifically, the licensee was committed to Regulatory Guide (RG) 1.9, Selection, Design, and Qualification of Diesel-Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants, Rev. 0, with an allowance to exceed 10% of nominal voltage for up to 60% of the load sequence interval, and the licensee inappropriately determined they could apply portions of RG 1.9, Rev. 1, instead of the specific deviation from Rev. 0 criteria that was previously accepted by the NRC without prior approval.

Significance/Severity: Green. Severity Level IV. The inspectors assessed the significance of the finding using SDP Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2 - Mitigating Systems Screening Questions, dated January 1, 2021. The inspectors determined that the finding was of very low safety significance (Green)as the finding was a deficiency affecting the design or qualification of the emergency diesel generator 2A-A, but the emergency diesel generator maintained its operability. In accordance with Section 6.1.d.2 of the NRC Enforcement Policy (dated January 15, 2020) this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance (i.e., green finding).

Corrective Action References: CR

EXIT MEETINGS AND DEBRIEFS

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

• On April 14, 2021, the inspectors presented the NRC inspection results to Mr. John Taylor - Senior Manager SQN Design Engineering and other members of the licensee staff.

• On March 12, 2021, the inspectors presented the Onsite Exit inspection results to Thomas Marshall - Site Vice President Sequoyah Nuclear Plant and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

71111.17T Calculations

1-FCV-62-091

Documentation of Design Basis Review, Required

Thrust Cale and Valve and Actuator Capability

Assessment for 1-FCV-62-091 (GL 89-10)

Rev. 6

31C-D053-EPM-RG-

060987

HVAC Equipment Requirements Evaluation: Turbine

Driven Auxiliary Feedwater Pump Room

Rev. 6

DS-M18.2.21

Motor Operated Valve Thrust and Torque Calculations

07/24/2020

ED00009992018000092

Turbine Driven Auxiliary Feedwater Dedication,

Qualification, and Software Verification & Validation

Documentation

Rev. 0

MDQ0000032015000278 Auxiliary Feedwater Pump Alternate Min-flow Relief

Valve Sizing

Rev. 0

N2-62-7A

Summary of Piping Analysis N2-62-7A

NDQ0063980038

RWST and Containment RHR Sump Safety and

Operational Limits, RWST Setpoint Required Accuracy

and LBLOCA, SBLOCA Sump Minimum Levels

Rev. 12

SQN-CPS-051

Circuit Protection Device Evaluation

Rev. 68

SQN-CPS-057

Vital Control Power System Loading Channel I And

Continuous Loading Evaluation of Protective Devices

in the 120v Ac Vital Instrument Power Boards

Rev. 103

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

Rev. 119

SQN-CPS-059

Vital Control Power System Loading Channel III and

Continuous Loading Evaluation of Protective Devices

in the 120v Ac Vital Instrument Power Boards

Rev. 103

SQN-CPS-060

Vital Control Power System Loading Channel IV and

Continuous Loading Evaluation of Protective Devices

in the 120v Ac Vital Instrument Power Boards

Rev. 109

SQN-E3-002

Diesel Generator Loading Analysis

Rev. 63

SQN-E3-011, App. K

Overvoltage Analysis

01/22/1988

SQN-EEB-ETR

Demonstrated Accuracy Calculation ETR

Rev. 2

SQN-MEB-2-FCV-67-

Documentation of Design Basis Review, Required

Rev. 1

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

091

Thrust Call and Valve & Actuator Capability

Assessment For 2-FCV-67-091

SQS20110

Emergency and Abnormal Operating Procedure

Setpoints

Rev. 30

TIECS55

Summary of Harsh Environment Conditions for

Sequoyah Nuclear Plant

Rev. 38

Corrective Action

Documents

0848636

1468371

1380153

283446

1084295

1084294

1076269

1076179

1064785

1064736

1049129

1464713

1192952

260178

0801415

1104481

1641119

Corrective Action

Documents

Resulting from

Inspection

CR 1677990

SQN 50.59 2021010 CRFSAR SQEP-EEB-88-018

Cannot be Found

03/11/2021

CR 1678167

NRC SQN 59.59 2021010 inspection

03/11/2021

Drawings

0-47W610-46-1

Mechanical Control Diagram Feedwater Control

System

04/29/2020

1, 2-45N765-6

Wiring Diagrams, 6900 Shutdown AUX power,

Schematic Diagram

Rev. 6

1,2-45W880-28A

Conduit & Grounding Penetration Sealing

Rev. 8

1,2-88405-2

Forged Bolted Bonnet Primary Nuclear Motor

Operated Gate Valve

Rev. 0

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

1-45N1635

Wiring Diagrams, Local Instrument Panels, Connection

Diagrams, SH 3

Rev. 4

1-47A941-73

Trust Requirements for Motor Operated Valve 1-FCV-

2-091

Rev. 4

2-45W2614-9

Wiring Diagrams Auxiliary Feedwater Pump & Turbine

Connection Diagrams

10/20/2020

2-45W646-6

Wiring Diagrams Feedwater Pump & Turbines

Schematic Diagrams

04/29/2020

Engineering

Changes

DCN 100096

NIS Time Delay Relay Project

Rev. 2

DCN 22346

Replace 480V Board Room Air Conditioners &

Associated Equipment (See attached)

Rev. A

DCN D23085

Replace breaker handle, breaker operating

mechanism and remove Kirk-Key Interlocks

Rev. A

DCN D23216

1,2-HS-062-0104C, -108C, -003-0118C, -0128C, -068-

0341AC, -0341DC, Admin. Change #04

DCN SQN-19-308

2A-A VORD Removal

Rev. 2

DCN SQN-19-856-B

Replace Obsolete EQ Transmitter 2-FT-3-168-

Rev. 1

DCN-D21892

Install CCUS 2-XC-43-55, -58, -61, AND -64 in the

Circuit Before Solenoid Valves 2-FSV-43-55, -58, -61,

and -64 to Regulate Current and Reduce Heat Rise

Rev. A

DEC 10075

B Train Shutdown Board Cleaning Maintenance

Crosstie

Rev. 0

DEC SQN-18-226

Limit Switch Variation for 2-FCV-67-91

Rev. 1

DEC SQN-19-699

EQ Binder revisions for SG Level Transmitter 1-LT-

003-0097-G

Rev. 0

EC 23813

Replace Steam Generator Blowdown Containment

Isolation Sample Valves 1- & 2-FSV-043-0055-B, -

0058-A, -0061-B, -0064-A

Rev. 5

ECP 100068

Motor Operated Potentiometers (MOPs) Replacement

for the Emergency Diesel Generators (EDGs)

Rev. 2

ECP 23085-02

Replace breaker handle, breaker operating

mechanism and remove Kirk-Key Interlocks

06/27/2019

ECP 23085-04

Replace breaker handle, breaker operating

mechanism and remove Kirk-Key Interlocks

06/25/2019

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

ECP SQN-19-785

ECP That Will Eliminate the Need for the Current SBO

Diesel Generators Creating the FLEX Program

Rev. 1

EDC E21195

ASCO has discontinued the production of 206-380 &

206-381

Rev. A

SQN-0-2015-082-001

Functionally Disabling VORD and DC amp meter for

Emergency Diesel 2A-A

Rev. 0

Miscellaneous

NRC Docket Nos. 50-327 and 328, Renewed Facility

Operating License Nos. DPR-77 and DPR-79,

Request to Modify Essential Raw Cooling Water Motor

Control Center Breakers and to Revise the Updated

Final Safety Analysis Report Sequoyah Nuclear Plant

Units 1 and 2 (SQN-TS-17-04)

03/09/2018

Sequoyah Nuclear Plant, Units 1 And 2-Issuance of

Amendments Re: Request To Modify Essential Raw

Cooling Water Motor Control Center Breakers And To

Revise The Updated Final Safety Analysis Report

(EPID L-2018-LLA-0060)

05/07/2019

White Paper, SQN Auxiliary Feedwater System

Feasibility Study for Min-flow Design Pressure

Increase

11/11/2015

Updated Final Safety Analysis Report, Chapter

10.4.7.2, Auxiliary Feedwater System

Amd. 30

Updated Final Safety Analysis Report, Section 15.4.1,

Major Reactor Coolant System Pipe Ruptures (Loss of

Coolant Accident)

Amd. 30

Sequoyah Nuclear Plant Units 1 and 2 - Nuclear

Steam Supply System Engineering Support Services -

Contract No. 2925 - Letter N10562

09/27/2010

EWR-18-DEC-067-615

Limitorque SMB-000 actuator

11/06/2018

EWR-18-DEE-067-616

Limitorque SMB-000 actuator

Rev. 1

EWR-18-DEM-067-617

Limitorque SMB-000 actuator

11/05/2018

S22 180116 001

Asco Controls, L.P Engineering Report. Rev.3 AQN-

18-226, Limit Switch Variation for 2-FCV-67-91

Rev. 1

SQN FSAR Section 9.3.4 Chemical & Volume Control System

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

SQN UFSAR Section

10.4.7.2

Auxiliary Feedwater System

SQNEQ-SOL-009

Target Rock Corporation Solenoid Valves

Rev. 31

Procedures

0-TI-OPS-000-004.0

Time Critical Operator Actions

Rev. 10

2-SI-EDC-202-220.A

Setpoint Verification and Calibration for Time Delay

Relays Associated with Automatic Load Sequence

Timers

Rev. 17

ES-1.3

Transfer to RHR Containment Sump

Rev. 24

NEDP-2

Design Calculation Process Control

Rev. 18

NEDP-8.0

Evaluations for Procurement of Materials, Items, and

Services

Rev. 7

NEDP-8.2

Technical Evaluation for Procurement of Safety

Related and Quality Related Materials and Items

Rev. 3

NEDP-8.4

Equivalency Evaluation for Procurement and

Replacement of Materials and Items

Rev. 7

NISP-IP-ENG-001

Standard Design Process (EB-17-06)

Rev. 1

NPG-SPP-06.9.3

Post-Modification Testing

Rev. 11

NPG-SPP-09.4

CFR 50.59 Evaluations of Changes, Tests, and

Experiments

Rev. 15

PMTI-100069-001

Turbine Driven Auxiliary Feedwater Governor Upgrade

Rev. 0

TI-28

Curve Book

Rev. 351

TVA-NQA-PLN89-A

Nuclear Quality Assurance Plan (NQAP) (Quality

Assurance Program Description)

Rev. 38

Work Orders

09-779231-000

2729648

115065173

115065174

115065175

115065176

117222626

118765154

119170828

115144038

20757820

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

2485666

20455241