IR 05000327/2019004
| ML20042C265 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 02/10/2020 |
| From: | Tom Stephen NRC/RGN-II/DRP/RPB5 |
| To: | Jim Barstow Tennessee Valley Authority |
| T. Stephen RGN-II/DRP | |
| References | |
| IR 2019001, IR 2019004 | |
| Download: ML20042C265 (29) | |
Text
February 10, 2020
SUBJECT:
SEQUOYAH, UNITS 1 AND 2 - INTEGRATED INSPECTION REPORT 05000327/2019004, 05000328/2019004 AND 07200034
Dear Mr. Barstow:
On December 31, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Sequoyah, Units 1 and 2. On January 22, 2020, the NRC inspectors discussed the results of this inspection with Mr. Scott Hunnewell and other members of your staff. The results of this inspection are documented in the enclosed report.
Two findings of very low safety significance (Green) are documented in this report. Two of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
A licensee-identified violation which was determined to be of very low safety significance 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 violations or the significance or severity of the violations 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, Units 1 and 2.
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, Units 1 and 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/
Thomas A. Stephen, Chief Reactor Projects Branch 5 Division of Reactor Projects
Docket Nos. 05000327, 05000328 and 07200034 License Nos. DPR-77 and DPR-79
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000327, 05000328 and 07200034
License Numbers:
Report Numbers:
05000327/2019004, 05000328/2019004 and 07200034/2019001
Enterprise Identifier: I-2019-004-0023
I-2019-001-0153
Licensee:
Tennessee Valley Authority
Facility:
Sequoyah, Units 1 and 2
Location:
Soddy Daisy, TN 37379
Inspection Dates:
October 01, 2019 to December 31, 2019
Inspectors:
A. Butcavage, Reactor Inspector
N. Childs, Resident Inspector
C. Dykes, Health Physicist
C. Fontana, Emergency Preparedness Inspector
D. Hardage, Senior Resident Inspector
D. Lanyi, Senior Operations Engineer
N. Morgan, Reactor Inspector
A. Nielsen, Senior Health Physicist
S. Sanchez, Senior Emergency Preparedness Inspector
Approved By:
Thomas A. Stephen, Chief
Reactor Projects Branch 5
Division of Reactor Projects
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Sequoyah, Units 1 and 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: 71114.0
List of Findings and Violations
Obstruction in Steam Generator Pressure Channel Sense Line Results in ESFAS Channel Inoperability Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000327/2019004-02 Open/Closed
[P.3] -
Resolution 71153 A self-revealed Green finding and associated non-cited Violation (NCV) of 10 CFR Part 50,
Appendix B, Criterion XVI, "Corrective Actions," was identified when the licensee failed to ensure the pressure instrumentation sense line for a steam generator pressure protection channel transmitter was not obstructed. Specifically, identified blockages in sense lines were not cleared, and resulted in an inoperable engineered safeguard feature actuation system (ESFAS) channel.
Failure to Verify Adequacy of ERCW Flow to CCS Train 2A After Implementing Temporary Modification Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000328/2019004-01 Open/Closed
[H.14] -
Conservative Bias 71153 A self-revealed Green finding and associated non-cited violation (NCV) of 10 CFR Part 50,
Appendix B, Criterion III, Design Control, was identified when the licensee failed to verify adequate essential raw cooling water (ERCW) system flow to train 2A component cooling system (CCS) components after implementing a temporary modification on the ERCW system.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000328/2019-001-00 LER 2019-001-00 for Sequoyah Nuclear Plant,
Unit 2, Component Cooling Water System Train A Inoperable Longer Than Allowed by Technical Specifications 71153 Closed LER 05000327/2019-003-00 LER 2019-003-00 for Sequoyah Nuclear Plant, 71153 Closed
Unit 1, Automatic Reactor Trip due to Negative Rate Trip as a Result of a Dropped Control Rod LER 05000327/2019-001-00 LER 2019-001-00 for Sequoyah Nuclear Plant,
Unit 1, Reactor Trip on Low-Low Steam Generator Level Due to the Loss of a Main Feedwater Pump.
71153 Closed
PLANT STATUS
Unit 1 began the inspection period at 76 percent rated thermal power (RTP) due to end of life coastdown conditions. On October 12, the unit was shutdown for refueling. The unit was returned to 100 percent RTP on November 30 and remained at or near 100 percent RTP for the remainder of the inspection period.
Unit 2 began the inspection period at 100 percent RTP. On October 24, an automatic turbine runback to 77 percent RTP occurred due to a reduction in #3 heater drain tank flow. The unit was returned to 100 percent RTP on October 27. On December 12, operators manually tripped the reactor following a loss of all #3 heater drain tank flow. The unit was returned to 100 percent RTP on December 16 and remained at or near 100 percent RTP for the remainder of the inspection period.
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 performed plant status activities described in IMC 2515, Appendix D, Plant Status, and conducted routine reviews using IP 71152, Problem Identification and Resolution. 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
71111.01 - Adverse Weather Protection
Seasonal Extreme Weather Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of seasonal cold temperatures in the following areas:
- Unit 1 and 2 refueling water storage tanks.
- Unit 1 and 2 condensate storage tanks.
71111.04Q - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (3 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 1 residual heat removal (RHR) system after core reload during U1R23.
- (2) Unit 1 RHR system after realignment to standby emergency core cooling system (ECCS) injection following Unit 1 outage.
- (3) Unit 2 auxiliary feedwater (AFW) system after realignment following forced outage.
71111.05A - Fire Protection (Annual)
Annual Inspection (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated fire brigade performance during an announced fire drill at the 2B 480V turbine building vent board in the railroad bay on November 13, 2019.
71111.05Q - Fire Protection
Quarterly Inspection (IP Section 03.01) (4 Samples)
The inspectors evaluated fire protection program implementation in the following selected areas:
- (1) Unit 1 reactor building - El. 679 and 701 on November 1, 2019.
- (2) Unit 1 and Unit 2 auxiliary building - El 653 on November 21, 2019.
- (3) Unit 1 and Unit 2 auxiliary building - El 669 on November 21, 2019.
- (4) Unit 1 and Unit 2 control building - El 669 on December 19, 2019.
71111.08P - Inservice Inspection Activities (PWR) PWR Inservice Inspection Activities Sample (IP Section 03.01)
- (1) The inspectors verified that the reactor coolant system boundary, reactor vessel internals, risk-significant piping system boundaries, and containment boundary are appropriately monitored for degradation and that repairs and replacements were appropriately fabricated, examined and accepted by reviewing the following activities from October 14 to October 18, 2019:
03.01.a - Nondestructive Examination and Welding Activities
- Visual Test (VT)-3 Pipe Support, Chemical and Volume Control System, (CVCS)-75, Seal Water Injection, Reactor Coolant Pump (RCP)-3, ASME Class 2 (Observed).
- VT-3 Pipe Support, CVCS-79, Seal Water Injection, RCP-3, ASME Class 2 (Observed).
- Liquid Penetrant Test (PT) of Integral Attachment Welds on Pipe Support Number 1-SIH-135, ASME Class 2 (Observed)
- Ultrasonic Test (UT) of Elbow-Nozzle Weld, Component ID, CVCF-214, Chemical and Volume Control System, ASME Class 2, (Observed).
- CR-1463750, U-2, R-22 Lower Reactor Head Penetration Rejectable Indication, Accepted for Continued Service, Notice of Indication (NOI) 2-SQ-455, Including Evaluation and Disposition, (Note this is a U-2 example provided as Typical Site Disposition Approach for a Notice Of Indication Determined to be Acceptable for Continued Operation).
- WO 118840720, Repair/Replacement Welds NDE, RHR Pump Cooling Coils and Associated Piping, UNID: SQN-1-CLR-030-0175 (Reviewed).
03.01.b - Pressurized-Water Reactor Vessel Upper Head Penetration Examination Activities
- Reactor Pressure Vessel Closure Head Visual Enhanced (VE), Control Rod Drive Mechanism (CRDM) Penetration Examination, Head Penetrations, (Pen)66, Location (L)-180, Scan Path (SP-1), Pen-74 L-180 SP-1, Pen-54 L-90 SP-2, Pen-23 L-90 SP-3, Pen-6 L-90 SP-4, Pen 21 L-90 SP-17, Pen-4 L-90 SP 7 (Reviewed).
- Work Order 119446734, Monitoring of Reactor Head Canopy Seal Welds for Leakage, Procedure 0-PI-DXX-068-100.R Completion (Reviewed).
03.01.c - Pressurized-Water Reactor Boric Acid Corrosion Control Activities
- The inspectors evaluated the licensees boric acid corrosion control program performance, through field walk-downs inside containment and review of the boric acid program corrective actions and evaluations of reported leakage as listed in the documents reviewed section. This review included corrective actions planed to address refueling cavity seal leakage presented in CR-
===1557752.
71111.11A - Licensed Operator Requalification Program and Licensed Operator Performance
Requalification Examination Results (IP Section 03.03)===
- (1) The licensee completed the annual requalification operating examinations required to be administered to all licensed operators in accordance with Title 10 of the Code of Federal Regulations 55.59(a)(2), "Requalification Requirements," of the NRC's "Operator's Licenses." During the week of December 23, 2019, the inspector performed an in-office review of the overall pass/fail results of the individual operating examinations and the crew simulator operating examinations in accordance with Inspection Procedure (IP) 71111.11, "Licensed Operator Requalification Program." These results were compared to the thresholds established in Section 3.02, "Requalification Examination Results," of IP 71111.11.
The inspectors reviewed and evaluated the licensed operator examination failure rates for the requalification annual operating examination completed on December 9, 2019.
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
(1)
- The inspectors observed and evaluated licensed operator performance in the main control room during the Unit 1 shutdown for refueling outage U1R23 on October 12, 2019.
- The inspectors observed and evaluated licensed operator performance in the main control room during the Unit 1 reactor startup on November 26, 2019.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated job performance measures performed on the simulator for the annual requalification operating test on November 14, 2019.
71111.12 - Maintenance Effectiveness
Routine Maintenance Effectiveness Inspection (IP Section 02.01) (1 Sample)
The inspectors evaluated the effectiveness of routine maintenance activities associated with the following equipment and/or safety significant functions:
- (1) Failure of 2-VLV-067-0551 to provide adequate essential raw cooling water (ERCW)flow to train 2A component cooling system (CCS) heat exchangers (CDE 3075) on August 28, 2019.
Quality Control (IP Section 02.02) (1 Sample)
The inspectors evaluated maintenance and quality control activities associated with the following equipment performance activities:
- (1) Commercial Grade Dedication Packages for Reactor Coolant Pump Motor Refurbishment (TVA Contract 12000, PO 4138823), dated October 2019.
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (4 Samples)
The inspectors evaluated the risk assessments for the following planned and emergent work activities:
- (1) Unit 1, Yellow shutdown risk week of October 14-18 while reactor coolant system (RCS) level was below the reactor flange level and during reactor core offload, including review of defense in depth protected equipment for U1R23.
- (2) Unit 1, Yellow shutdown risk week of October 26-30 during reactor core reload and while the RCS level was below the reactor flange level for reactor reassembly, including review of defense in depth and protected equipment for U1R23.
- (3) Unit 1, Yellow shutdown risk week of November 16-22 during core shuffle activities and with RCS level below the flange level for reactor reassembly, including review of defense in depth and protected equipment while the licensee was waiting on approval of a license amendment request to remove control rod H-8.
- (4) Unit 1 and Unit 2, week of December 15-21, including protected equipment status reviews for scheduled maintenance on the Unit 2 turbine driven auxiliary feedwater pump, cleaning of the 2A1 CCS heat exchanger and emergent loss of the 1B-B 6.9kV shutdown board.
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 02.02) (4 Samples)
The inspectors evaluated the following operability determinations and functionality assessments:
(1)1A1 CCS HX has two ERCW outlet piping pinhole leaks (CR1551560) on September 23, 2019.
- (2) SQN calculation EDQ0002022016000329 issues (CR 1555475) on October 8, 2019.
- (3) Failed seat leak check of 1-FCV-74-3 (CR 1560035) on October 24, 2019.
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)
The inspectors evaluated the following temporary or permanent modifications:
- (1) DCN 22643 Stage 6, Install Target Rock PORV 1-68-334B.
71111.19 - Post-Maintenance Testing
Post-Maintenance Test Sample (IP Section 03.01) (4 Samples)
The inspectors evaluated the following post maintenance tests:
- (1) WO 118929705, 1-VLV-63-553 Cold Leg SI Check Valve Seat Leak Repair, on November 2, 2019.
- (2) WO 119855374, 1-PCV-68-340 Replace PORV 340, on November 5, 2019.
- (3) PMTI-22703-01/02, Post Modification Testing for Turbine Driven Auxiliary Feedwater Governor Upgrade, on November 3, 2019.
- (4) WO 120906807, Repack SIS CCP Injection Tank Inlet MOV, on November 23, 2019.
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated Unit 1 refueling outage (U1R23) activities from October 12, 2019, to November 27, 2019.
71111.22 - Surveillance Testing
The inspectors evaluated the following surveillance tests:
Surveillance Tests (other) (IP Section 03.01)
(1)1-SI-SXV-063-206.0, Residual Heat Removal Cold Leg Primary and Secondary Check Valve Integrity Test, on November 3, 2019.
(2)1-SI-SXP-003-201.S, Turbine Driven Auxiliary Feedwater Pump 1A-S Performance Test, on November 5, 2019.
Inservice Testing (IP Section 03.01) (1 Sample)
(1)1-SI-SXP-003-202.A, Motor Driving Auxiliary Feedwater Pump 1A-A Comprehensive Performance Test, on October 9, 2019.
RCS Leakage Detection Testing (IP Section 03.01) (1 Sample)
(1)0-SI-OPS-068-137.0, Reactor Coolant System Water Inventory - U1, on December 5, 2019.
Containment Isolation Valve Testing (IP Section 03.01) (1 Sample)
(1)0-SI-SLT-043-258.1, Containment Isolation Valve Local Leak Rate Test Post Accident Sampling, on October 23, 2019 (X-103).
Ice Condenser Testing (IP Section 03.01) (2 Samples)
(1)1-SI-MIN-061-106.0, Ice Condenser - Flow Passage Inspection, on October 31, 2019.
(2)0-SI-MIN-061-109.0, Ice Condenser Intermediate and Lower Inlet Doors and Vent Curtains, on November 1, 2019.
FLEX Testing (IP Section 03.02) (1 Sample)
(1)0-PI-SFT-360-001.0, Flex Pumps Operating Instructions for Maintenance Activities, on August 27, 2019.
71114.02 - Alert and Notification System Testing
Inspection Review (IP Section 02.01-02.04) (1 Sample)
- (1) The inspectors evaluated the maintenance and testing of the alert and notification system during the week of December 9, 2019.
71114.03 - Emergency Response Organization Staffing and Augmentation System
Inspection Review (IP Section 02.01-02.02) (1 Sample)
- (1) The inspectors evaluated the readiness of the Emergency Response Organization during the week of December 9, 2019.
71114.04 - Emergency Action Level and Emergency Plan Changes
Inspection Review (IP Section 02.01-02.03) (1 Sample)
- (1) The inspectors evaluated submitted Emergency Action Level, Emergency Plan, and Emergency Plan Implementing Procedure changes during the week of December 9, 2019. This evaluation does not constitute NRC approval.
71114.05 - Maintenance of Emergency Preparedness
Inspection Review (IP Section 02.01 - 02.11) (1 Sample)
- (1) The inspectors evaluated the maintenance of the Emergency Preparedness Program during the week of December 9,
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls
Radiological Hazard Assessment (IP Section 02.01) (1 Sample)
The inspectors evaluated radiological hazards assessments and controls.
- (1) The inspectors reviewed the following:
Radiological surveys SQN-M-20190927-2, 690' General Area, 09/27/2019 SQN-M-20191009-17, 690' General Area, 10/09/2019 SQN-M-20191022-21, A211 RHR Pump Room 1A-A, 10/22/2019 18:30 SQN-M-20191022-6, A211 RHR Pump Room 1A-A, 10/22/2019 03:46 SQN-M-20191016-2, R135 U1R23LC PRT Update, 10/16/2019 SQN-M-20191020-12, R135 U1R23LC PRT Cut-out 68-552 & 68-555, 10/20/2019 SQN-M-20191021-6, 68-552/555 Prep and decon, 10/21/2019
Air sample survey records NISP-RP-003-A1, U1 Aux 6531A A RHR, RWP 19110032 Survey # 102219005, 10/22/19 NISP-RP-003-A1, U1 1A RHR Pump Room Packing Removal, RWP 19110032 Survey # 102219002, 10/22/19 NISP-RP-003-A1, 653 1A-A RHR Pp Rm, RWP 19110032 Survey # 102219011, 10/22/19 NISP-RP-003-A1, U1 R23 LC PZR, RWP 1200072 Survey # 102019010,10/20/2019
Instructions to Workers (IP Section 02.02) (1 Sample)
The inspectors evaluated instructions to workers including, labels, radiation work permits and electronic dosimeter alarm setpoints used to access high radiation areas.
- (1) The inspectors reviewed the following:
Radiation work permits (RWP), including RWPs for airborne areas if available RWP 19000033, Auxiliary Building-all areas, Radwaste-Locked High Radiation Area-routine activities, Revision 4 RWP 19140012, Upper Containment-all areas, U1R23 Upper Ctmt/ 734' GA/ SFP/
Top of Pressurizer-HRA/HCA, Revision 0 RWP 19120012, Lower Containment-all areas, U1R23 Lower Ctmt-HRA elevated dose rate work, Revision 0 RWP 19140083, Upper Containment-all areas, U1R23 Upper Ctmt/ 734' Equipment Zone/ 734' GA/ SFP-HRA H8 Drive Shaft Removal and Replacement, Revision 0 RWP 19110033, Auxiliary Building-all areas, U-1 Aux Bldg & Annulus to include satellite RCA's in association with valve work/ crud burst/ Rx Cavity drain down and clean up-Locked HRA/HCA areas, Revision 0 RWP 19110032, Auxiliary Building-all areas, U-1 Aux Bldg & Annulus to include satellite RCA's in association with valve work/ crud burst/ Rx Cavity drain down and clean up-HRA/HCA areas, Revision 0 RWP 19120072, Lower Containment-all areas, U1R23-Lwr Ctmt-High Radiation Area-HCA work allowed. 552/555 Valve replacement/ setup and demobe, Revision 1
Electronic alarming dosimeter alarms 01/08/2019 RWP #19201111 Dose Rate alarm-71 setpoint/ 89 actual No other unexpected alarms occurred during this inspection period
Labeling of containers Truck bay Dry active waste building Waste packaging area
Contamination and Radioactive Material Control (IP Section 02.03) (1 Sample)
The inspectors evaluated licensee processes for monitoring and controlling contamination and radioactive material. The inspectors verified transactions of nationally tracked sources had been reported.
- (1) The inspectors verified the following sealed sources are accounted for:
- 662, Am-241/Be-10
- 1207, Cs-137
- 1208, Cs-137
- 883, Cs-137
- 775, Ba-137/Cs-137
Radiological Hazards Control and Work Coverage (IP Section 02.04) (1 Sample)
The inspectors evaluated in-plant radiological conditions during facility walkdowns and observation of radiological work activities
- (1) The inspectors also reviewed and observed the following risk significant radiological work activities:
H8 drive shaft replacement A211 RHR pump room 1A-A 74-3 valve work R135 U1R23 pressurizer work
High Radiation Area and Very High Radiation Area Controls (IP Section 02.05) (1 Sample)
- (1) The inspectors evaluated risk-significant high radiation area and very high radiation area controls, including postings and physical controls.
Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section 02.06) (1 Sample)
- (1) The inspectors evaluated radiation worker awareness and performance and radiation protection technician proficiency.
71124.08 - Radioactive Solid Waste Processing and Radioactive Material Handling, Storage,
and Transportation
Radioactive Material Storage (IP Section 02.01)
The inspectors evaluated radioactive material storage areas for compliance with postings, labels, radiological controls and physical controls during plant walkdowns. Inspectors also evaluated the material condition of several containers of radioactive materials for signs of degradation. This included the following areas:
- (1) Dry active waste (DAW) warehouse, waste packaging area, and refuel floor.
Radioactive Waste System Walkdown (IP Section 02.02) (1 Sample)
The inspectors evaluated liquid and solid radioactive waste processing systems during plant walkdowns for material condition and alignment with descriptions in the UFSAR and Process Control Program (PCP). The inspectors also evaluated waste stream mixing methodologies and changes to the systems. This included the following systems:
- (1) Liquid waste processing skid, wet solid waste dewatering, hot trash handling and storage, abandoned boric acid evaporators.
Waste Characterization and Classification (IP Section 02.03) (1 Sample)
The inspectors evaluated the radioactive waste characterization and classification for the following waste streams:
- (1) DAW 11/15/18 and radwaste resin 8/26/18.
Shipment Preparation (IP Section 02.04) (1 Sample)
The inspectors evaluated and observed the following radioactive material shipment preparation processes:
- (1) Shipment of DAW disposable protective clothing, 10/23/19.
Shipping Records (IP Section 02.05) (1 Sample)
The inspectors evaluated the following non-excepted package shipment records:
- (1) Shipment 18-0703, Type B, Filters Shipment 19-0401, Type A, Filters Shipment 19-0506, Low Specific Activity (LSA), DAW Shipment 19-0604, Type B, Specimen Debris Shipment 19-0701, Type B, Resin
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification
The inspectors verified licensee performance indicators submittals listed below:
EP01: Drill/Exercise Performance (IP Section 02.12)===
- (1) EP01: Drill & Exercise Performance for the period July 1, 2018, through September 30, 2019.
EP02: ERO Drill Participation (IP Section 02.13) (1 Sample)
- (1) EP02: Emergency Response Organization Drill Participation for the period July 1, 2018, through September 30, 2019.
EP03: Alert & Notification System Reliability (IP Section 02.14) (1 Sample)
- (1) EP03: Alert & Notification System Reliability for the period July 1, 2018, through September 30, 2019.
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
- (1) December 2018-September 2019.
71152 - Problem Identification and Resolution
Semiannual Trend Review (IP Section 02.02) (1 Sample)
- (1) The inspectors reviewed the licensees corrective action program for potential adverse trends in the areas of human performance and equipment performance that might be indicative of a more significant safety issue.
Annual Follow-up of Selected Issues (IP Section 02.03) (1 Sample)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) The inspectors conducted safety culture interviews with licensing, performance improvement, and emergency preparedness personnel. The inspectors concluded the interviewees understood the various methods of reporting safety concerns and were willing to use them without fear of retaliation.
71153 - Followup of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 05000327/2019-001-00, Reactor Trip on Low-Low Steam Generator Level Due to the Loss of a Main Feedwater Pump (ADAMS accession: ML19163A049). The inspectors determined that it was not reasonable to foresee or correct the cause discussed in the LER, therefore no performance deficiency was identified. The inspectors also concluded that no violation of NRC requirements occurred.
- (2) LER 05000327/2019-003-00, Automatic Reactor Trip due to Negative Rate Trip as a Result of a Dropped Control Rod (ADAMS accession: ML19297G404). The circumstances surrounding this LER are documented in the Results section.
- (3) LER 05000328/2019-001-00, Component Cooling Water System Train A Inoperable Longer Than Allowed by Technical Specifications (ADAMS accession:
ML19322A627). The circumstances surrounding this LER are documented in the Results section.
Personnel Performance (IP Section 03.03) (2 Samples)
- (1) The inspectors evaluated the licensees response to an automatic turbine runback to 77 percent RTP on Unit 2, which occurred due to a reduction in #3 heater drain tank flow on October 24, 2019.
- (2) The inspectors evaluated the Unit 2 manual reactor trip and licensees performance on December 12,
OTHER ACTIVITIES
- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL
60855.1 - Operation of an Independent Spent Fuel Storage Installation at Operating Plants Operation of an Independent Spent Fuel Storage Installation at Operating Plants
- (1) The inspectors evaluated the licensees activities related to long-term operation and monitoring of their independent spent fuel storage installation.
INSPECTION RESULTS
Licensee-Identified Non-Cited Violation 71114.05 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: From April 2004 to June 2018, the licensee failed to maintain the effectiveness of its emergency plan when portions of the main steam line (MSL) radiation monitor release rate correction factors (particularly steam flow and mass) applied to the Integrated Computer System (ICS) input were incorrectly implemented. Using incorrect MSL radiation monitor data obtained from the ICS could lead to inaccurate dose assessment. This issue was determined by the inspectors to not be an immediate safety concern because it was corrected in June 2018, and the appropriate correction factors are now being used. The issue was placed into the licensees corrective action program as condition report (CR) 1498586 and an extent of condition review, along with a root cause evaluation, was performed. These documents were reviewed by the inspectors.
Contrary to 10 CFR 50.54(q)(2) and 10 CFR 50.54(b)(9), Sequoyah Nuclear Station (SQN)failed to maintain the effectiveness of its emergency plan by not ensuring the MSL radiation monitor release rate correction factors remained correct, thereby affecting the licensees ability to provide technically accurate dose assessments.
Significance/Severity: Green.
Corrective Action References: The licensee implemented several corrective actions to address the issue with the MSL radiation monitor release rate correction factors. Those actions were documented in CR 1498586, CR 1191019, and CR 1515689.
Obstruction in Steam Generator Pressure Channel Sense Line Results in ESFAS Channel Inoperability Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems
Green NCV 05000327/2019004-02 Open/Closed
[P.3] -
Resolution 71153 A self-revealed Green finding and associated non-cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified when the licensee failed to ensure the pressure instrumentation sense line for a steam generator pressure protection channel transmitter was not obstructed. Specifically, identified blockages in sense lines were not cleared, and resulted in an inoperable engineered safeguard feature actuation system (ESFAS) channel.
Description:
On August 27, 2019, Unit 1 control rod H-8 dropped into the core. The Unit 1 reactor subsequently tripped on a negative startup rate trip signal. Following the trip, steam generator (SG) loop 3 pressure transmitter, 1-PT-1-23, demonstrated sluggish behavior. The sluggish response could have challenged the engineered safety features actuation system (ESFAS) input function associated with SG pressure. 1-PT-1-23 is one of three pressure transmitters used to monitor SG loop 3 pressure. These three transmitters are used, with their instrument loops, to provide two-out-of-three logic for steam generator pressure required by Technical Specifications for ESFAS. Due to this observed sluggish response, operators declared 1-PT-1-23 inoperable.
The sluggish response was caused by the sensing line to the pressure transmitter being plugged. Plugging of the steam transmitter sense lines has been previously identified at Sequoyah. Following the April 14, 2019 Unit 1 reactor trip, the licensee noted that 1-PT-1-23 was lagging the indication of other steam generator pressure transmitters, and declared the channel inoperable. 1-PT-1-23 was removed from service and the sense line was rodded out to remove accumulated material. 1-PT-1-23 was then returned to operable status. Following the August reactor trip, the licensee determined the maintenance performed in April did not correct the condition and 1-PT-1-23 was still inoperable.
Corrective Actions: Work orders were completed to backfill and blow out the sensing lines prior to startup in August 2019. Additionally, the sensing lines were cut out and replaced during the 1R23 refueling outage in October 2019.
Corrective Action References: CR 1556233, CR 1544227, CR 1507948, CR 915880, WO 120407126, WO 120755617, WO 120710728, WO 120766170, PER 915889-003
Performance Assessment:
Performance Deficiency: The licensees failure to ensure previously identified blockages in the pressure instrumentation sense line for a steam generator pressure protection channel transmitter were removed was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance 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 identified blockage in sense line resulted in an inoperable ESFAS channel.
Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The finding screened to Green because all four questions in Exhibit 2, "Mitigating Systems Screening Questions" were answered "no".
Cross-Cutting Aspect: P.3 - Resolution: The organization takes effective corrective actions to address issues in a timely manner commensurate with their safety significance. The licensee created corrective action work orders several years prior to the event to clear the steam generator pressure transmitter sense lines. However, delays in scheduling and ineffective clearing of the lines when maintenance was performed resulted in continued inoperability of the pressure transmitter.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criteria XVI, Corrective Actions, required, in part, that measures be established to assure conditions adverse to quality are promptly identified and corrected. Contrary to the above, from April 14 until August 27, 2019, the measures established by the licensee failed to correct an identified condition adverse to quality. Specifically, the licensee identified a condition adverse to quality that resulted in the inoperability of an ESFAS channel but did not correct the condition.
Enforcement Action: This violation is being treated as an non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Verify Adequacy of ERCW Flow to CCS Train 2A After Implementing Temporary Modification Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems
Green NCV 05000328/2019004-01 Open/Closed
[H.14] -
Conservative Bias 71153 A self-revealed Green finding and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified when the licensee failed to verify adequate essential raw cooling water (ERCW) system flow to train 2A component cooling system (CCS) components after implementing a temporary modification on the ERCW system.
Description:
On June 17, 2019, the licensee implemented temporary modification SQN-2-2018-067-001 on Unit 2, to swap the function of 2-FCV-67-146 with 2-VLV-67-551. These valves regulate ERCW flow through train 2A CCS heat exchangers. 2-FCV-67-146 has two pre-determined throttled positions (35% and 50% open) which is normally controlled from the main control room. This valve was determined to be degraded, so the temporary modification was implemented to de-energize 2-FCV-67-146 in a throttled open position and use manual valve 2-VLV-67-551 to throttle ERCW flow. The 35% position is to ensure desired ERCW flows to Unit 1 loads during an accident on that unit, which in turn reduces flows to Unit 2 loads, but is still adequate for normal Unit 2 operation. The 50% position is to ensure desired ERCW flow to Unit 2 loads during an accident on that Unit, which in turn reduces ERCW flow to Unit 1 loads, but is still adequate for normal Unit 1 operation.
On August 24, 2019, while performing an ERCW flush of Unit 1 CCS heat exchangers with 2-VLV-67-551 in the 35% throttled position, unexpected CCS alarms were noted in the main control room along with rising CCS heat exchanger outlet temperatures. The flushing activity was aborted and investigation into the cause of the unexpected conditions began. On August 28, 2019, a review of ERCW pump Section XI test data revealed that the 35% and 50%
throttle positions on 2-VLV-67-551 did not provide the expected ERCW flows. Operations subsequently declared ERCW inoperable and entered TS LCO 3.7.8, Condition B. An ERCW flow test was performed on August 28, 2019 to determine the correct throttle positions for 2-VLV-67-551, and with the system restored to appropriate alignment, LCO 3.7.8, Condition B was exited.
A past operability evaluation, completed on September 20, 2019, determined that the event rendered train 2A CCS inoperable, due to insufficient ERCW flow, from June 17, 2019 to August 28, 2019 (72 days), which is longer than allowed by TS 3.7.7 (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />).
Corrective Actions: An ERCW flow balance test was completed on August 28, 2019 to determine the correct throttle positions for 2-VLV-67-551 and restore system operability. The temporary modification will be removed during the Unit 2 refueling outage when 2-FCV-67-146 is repaired.
Corrective Action References: CR 1544846
Performance Assessment:
Performance Deficiency: The failure to verify adequate ERCW flow to train 2A CCS components after implementing temporary modification SQN-2-2018-067-001 was a performance deficiency. Specifically, the licensee failed to perform any post modification testing to verify the correct throttle positions for 2-VLV-67-551 and corresponding ERCW required flow rates.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control 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. The performance deficiency adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of CCS train 2A to respond to initiating events and prevent undesirable consequences.
Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The issue screened as requiring a detailed risk evaluation because the finding represented an actual loss of function of CCS train 2A for greater than its Technical Specifications allowed outage time.
A detailed risk evaluation was performed by a regional senior risk analyst using SAPHIRE Version 8.2.1 and NRC Sequoyah SPAR model Version 8.50. Since the Unit 2 valve would have been manually repositioned by plant operators following a safety injection system actuation for either unit, initiating events for both units were considered separately and then summed to estimate the cumulative impact on core damage frequency for Unit 2. A total exposure time of 1735 hours0.0201 days <br />0.482 hours <br />0.00287 weeks <br />6.601675e-4 months <br /> was used for the analysis.
Unit 1 initiators: The analyst performed an assessment for all Unit 1 accident sequences which result in a safety injection and multiplied those by the exposure time. The analyst assumed these initiators would be accompanied by a transient on Unit 2 with a complete loss of the Unit 2 A-train CCS heat exchanger heat removal function due to the performance deficiency.
Unit 2 initiators: The analyst adjusted the SPAR model to account for the manual valve throttling action and the availability of the Unit 2 containment spray (CS) system to remove heat. The analyst conservatively assumed any human error repositioning the valve would result in loss of the Unit 2 A-train CCS heat exchanger heat removal function during the recirculation phase. The analyst performed a condition assessment for all internal Unit 2 accident sequences.
The dominant sequences involved the failure to establish high pressure recirculation following a small loss of coolant accident initiating event. The risk was mitigated by the availability of the Unit 2 A-train CS system to remove heat during the recirculation phase. The analysis determined that the performance deficiency resulted in an increase in core damage frequency of <1E-6/year, a Green finding of very low safety significance.
Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe to proceed, rather than unsafe in order to stop. An ERCW flow verification was not conducted after the temporary modification was installed. The licensee assumed that the flow characteristics were the same for 2-VLV-67-551 and 2-FCV-067-0146-A and calculated the throttle positions for 2-VLV-67-551 based on the number of valve turns and opted not to perform testing to verify actual ERCW flow.
Enforcement:
Violation: 10 CFR Part 50, Appendix B, Criterion III, Design Control, required, in part, that design control measures shall provide for verifying or checking the adequacy of design. Contrary to the above, on June 17, 2019, the licensee implemented temporary modification SQN-2-2018-067-001 but did not perform post modification testing to verify adequate ERCW flow to CCS components.
Enforcement Action: This violation is being treated as an non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Minor Violation 71153 Minor Violation: On August 28, 2019, the licensee became aware that a recently modified ERCW flow control valve was not capable of providing adequate ERCW flow to its intended loads, which included Unit 2 train A CCS (reference the associated Green NCV documented in the Results section of this report for additional details). LER 05000328/2019-001-00, Component Cooling Water System Train A Inoperable Longer Than Allowed by Technical Specifications, was submitted to the NRC 79 days later, on November 15, 2019. The inspectors engaged the licensee regarding time of discovery and basis for the LER submittal date and determined that the LER should have been submitted by October 27, 2019, based on a discovery date of August 28, 2019; the licensee was aware on August 28, 2019, that the condition impacted CCS and had also encountered CCS alarms and temperature anomalies on August 24, 2019, when the subject valve was manipulated.
10 CFR 50.73(a)(1) required, in part, that a licensee shall submit a licensee event report (LER) for any event of the type described in 10 CFR 50.73 (a)(2) within 60 days of discovery of the event. Contrary to the above, on November 15, 2019, the licensee submitted LER 05000328/2019-001-00 to report the Unit 2 train A CCS loss of safety function and inoperability longer than allowed by Technical Specifications, but failed to submit the LER within 60 days of the August 28, 2019 event discovery date.
Screening: The inspectors determined the performance deficiency was minor. The inspectors reviewed NRCs Enforcement Policy, dated May 28, 2019, and determined that the licensees failure to report within 60 days of discovery was consistent with a Severity Level IV violation. However, Section 2.2.1, Factors Affecting Assessment of Violations, of the Enforcement Policy further states that the severity level of an untimely report, in contrast to no report, may be reduced depending on the circumstances. In this case, system operability was restored on August 28, 2019 when the licensee corrected the valve position and restored adequate ERCW flow. Therefore, the inspectors determined that the untimely report did not adversely impact the safety significance of the issue or impede the regulatory process and was a minor violation of 10 CFR 50.73(a)(1).
Enforcement:
This failure to comply with 10CFR50.73(a)(1) constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On January 22, 2020, the inspectors presented the integrated inspection results to Mr.
Scott Hunnewell and other members of the licensee staff.
- On October 25, 2019, the inspectors presented the RP baseline inspection results to Mr.
Matt Rasmussen, Site Vice President, and other members of the licensee staff.
- On December 13, 2019, the inspectors presented the Emergency Preparedness Program inspection results to Mr. Matt Rasmussen, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
60855.1
Procedures
0-SI-OPS-079-
001.0
Spent Fuel Storage Log
Rev. 18
Procedures
0-PI-0PS-000-
006.0
Freeze Protection
Rev. 64
0-VI-CEM-050-
2.3
ERCW Corrosion Inhibitor Injection Skid Operation
Rev. 12
1-PI-EFT-234-
706.0
Freeze Protection Heat Tracing Functional Test
Rev. 46
M&AI-27
Freeze Protection
Rev. 16
71111.04Q Procedures
0-SO-74-1
Residual Heat Removal System
Rev. 106
2-SO-3-2
Auxiliary Feedwater System
Revision 51
Fire Plans
TUR-0-706-02
Pre-Fire Plan - Unit 2 Side Elevation 706 of Turbine Building
Revision 7
Miscellaneous
Fire Drill Evaluation Report - Turbine Railroad Bay
11/13/2019
71111.05Q Fire Plans
AUX-0-653-00
Pre-Fire Plan - Auxiliary Building, El. 653
Revision 8
AUX-0-669-01
Pre-Fire Plan - Auxiliary Building, El. 669 (Unit 1 Side)
Revision 7
AUX-0-669-02
Pre-Fire Plan - Auxiliary Building, El. 669 (Unit 2 Side)
Revision 8
CON-0-669-00
Pre-Fire Plan - Control Building, El. 669
Revision 4
RXB-0-679-01
Pre-Fire Plan - Reactor Building, El. 679
Revision 3
RXB-0-701-01
Pre-Fire Plan - Reactor Building Annulus Area, El. 701 and
21
Revision 3
Corrective Action
Documents
CR-1411832 Initial Leak Assessment for Component 1-VLV-063-0868,
Residual Heat Removal Hot Leg, Emergency Core Cooling
System Vent
5/5/18
CR-1463750
U2-R22 Lower Reactor Head Penetration, Rejectable
Indication
11/6/18
Corrective Action
Documents
Resulting from
Inspection
CR-1557520
Dry Boric Acid at Packing For SQN-1-VLV-068-0057
10/16/19
CR-1557752
Loop 1 and 2 RCS Penetrations at Bio wall get correct
title>>> Requested from Scott on 12 /10/19, plus up date on
what actions taken to find leaks?
Drawings
30616-1050
173" PWR Vessel, Westinghouse, Section A-A, Weld Safe
End Material Detail
Rev. "G"
48N42
Sequoyah Units 1 & 2, Structural Steel Equipment Supports,
Rev. 4
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Reactor Supports, Elev. A-A
CCD No :1-1-
H20-0135-02
Safety Injection System Pipe Supports
Rev. 0
CCD NO. 1-1-
H34-0075-01
Chemical and Volume Control System Pipe Supports
Rev. 0
CCD No. 1-1-
H34-0079-01
Chemical And Volume Control System Pipe Supports
Rev. 0
CCD No: 1-1-
H20-0135-01 and
Safety Injection System Pipe Supports
Rev. 0
CHM-2434-C-03
- Q. Unit-1, Seal Water Injection, Chemical and Volume
Control System, Reactor Coolant Pump No. 3
Rev. 6
ISI-0448-C-10
SEQ U-1, High Pressure Safety Injection System Support
Locations
Rev. 2
ISI-0504-C-04
Sequoyah Nuclear Plant, Unit-1, Reactor Vessel Support
Locations
Rev.1
PE-D5942D
Sequoyah Unit-1, MK-3, MK-4, and MK-5, Cavity Opening
System
Rev. 1
SQN-1-CLR-030-
0175
Work Order 118840720, Cat 1 Weld Map, WMCR 2430,
Reference Weld Map 0-ERCW-15
8/14/19
Engineering
Evaluations
PER 1428361
Boric Acid Leakage Evaluation (Ref. Work Order 119695655) 1B-B Pump Casing Vent Line, UNID: 1-VLV-
2-0519
8/10/18
PER 1438325
Boric Acid leakage Evaluation (Ref. Work Order 117370142)
Spent Fuel Pit Pump "B" UNID: 0-PMP-078-0009B
8/13/18
PER 1483906
Boric Acid leakage Evaluation (Ref. Work Order 120178012)
Containment Spray Pump "1B", UNID: SGN-1-PMP-072-
0010
9/24/19
Miscellaneous
0-PI-DXI-000-
114.4
ASME SECTION XI ISI/NDE PROGRAM UNIT 1 AND UNIT
Rev.0001
0-SI-SXI-000-
201.0
ASME Section XI In-service Pressure Test
Rev. 0025
EIN - 100317156
Certificate of Method Qualification, Penetrant Testing,
Ultrasonic Testing, Visual Testing (VT-1) and (VT-3)
8/14/18
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
EIN -
BF2QSCVPQ2
Visual Acuity Examination Record for NDE Personnel
7/30/19
EIN -
BF2SCVPQ2
Personal Certificate of Method Qualification, Magnetic
Particle, Penetrant Testing, Ultrasonic Testing, Visual
Testing (VT-1) and (VT-3)
6/27/17
EIN - RGVIVT3PI
Visual Acuity Examination Record for NDE Personnel
8/2/19
EIN - RGVIVT3PI
Personal Certificate of Method Qualification, Visual Weld (
VT-W), Penetrant Testing, Ultrasonic Testing, Visual Testing
(VT-1) (VT-2) and (VT-3)
8/16/17
EIN 100317156
Visual Acuity Examination Record for NDE Personnel
7/10/19
EIN: JAJ4574
IHI Southwest Technologies Inc, Certificate of Qualification,
9/19/19
NIS-2
Repair Replacement Certification for Repair Replacement
Plan 118840720
6/11/19
Repair Replacement Activity Planning Form
2/1/19
NDE Reports
NOI-2-SQ-455
Notice of Indication of Reactor Vessel Lower Head,
Evidence of Leakage, Accumulated Residue, Discoloration.
All Conditions Were Not Present During Last Examination
Conducted In RFO-20, with Disposition Documents Attached
11/10/18
R-0336
Sequoyah Unit-1, Component ID: SCV-SCV-3, Steel
Containment Vessel, Visual Examination of IWE Surfaces
IAW N-VT-25
R-0338
Component ID, 1-CVCH-079, Rigid Support, VT-3 Visual
Examination Report, IAW Procedure N-VT-1
10/15/19
R-0345
Component ID, 1-SIH-135-1A, Integral Attachment Welds,
Penetrant Examination Report, IAW Procedure N-PT-9,
10/16/19
R-0346
Component ID, 1-CVCH-075, Rigid Support, VT-3 Visual
Examination Report, IAW Procedure N-VT-1
10/16/19
R-0348
Component ID, 1-SIH-135, Rigid Support, VT-3 Visual
Examination Report, IAW Procedure N-VT-1,
10/16/19
R-0372
Component ID, CVCF-214, Elbow-Nozzle Weld, Ultrasonic
Examination (UT) Report, IAW Procedure PDI-UT-2/N-UT-
R-0377
SQN Unit1, Cycle 23, Reactor Pressure Vessel Closure
Head Visual Enhanced (VE) Control Rod Drive Mechanism
10/24/19
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Penetration Examination, IAW N-VT-17, Final Report
R0041
Record of Visual Examination, Component ID, RVH-1
5/1/15
Weld No. 0-ER-
2799K
WO 118840720, ASME Class 3, Record of Liquid Penetrant
Examination
2/15/19
Weld No. 0-ER-
2799K
WO 118840720, ASME Class 3, Record of Post Weld VT-3
Visual Examination Report
3/14/19
Weld No. 0-ER-
2799M
WO 118840720, ASME Class 3, Record of Post Weld VT-3
Visual Examination Report
2/15/19
Weld No. 0-ER-
34094
WO 118840720, ASME Class 3, Record of Liquid Penetrant
Examination
2/19/19
Weld No. 0-ER-
34094
WO 118840720, ASME Class 3, Record of Post Weld VT-3
Visual Examination Report
2/19/19
Weld Number 0-
ER-2799M
WO 118840720, ASME Class 3, Record of Liquid Penetrant
Examination
2/15/19
ASME Section XI In-service Pressure Test, Record of VT-2
Leakage Examination, IAW, 0-SI-SXI-000-201.0
2/20/19
Procedures
0-SI-SXI-000-
201.0
ASME Section XI Pressure Test
Rev. 25
N-VT-17
Visual Examination For Leakage of PWR Reactor Head
Rev. 0010
PDI-UT-2
Generic Procedure for the Ultrasonic Examination of
Austenitic Pipe Welds
10/4/17
Work Orders
U1R23, Outage Chemical and Volume Control System
(CVCS) In-Service Examinations. (NRC Observation CVCS
Supports 75 and 79)
10/14/19
Replace the RHR Pump Cooler
2/12/19
ISI/CM/MISC-63/Safety Injection System Examinations (063) 10/15/19
Completion of Procedure 0-PI-DXX-068-100.R, Monitoring
of Reactor Head Canopy Seal Welds for Leakage
10/15/19
71111.11Q Miscellaneous
JPM-0-SO-202-
4(01)
Transfer 1A-A 6.9kv SD Board from Alternate to Normal
Supply
11/16/2017
JPM-EA-68-4(02)
Perform Emergency Boration per EA-68-4
11/16/2017
Procedures
0-GO-6
Power Reduction from 30% Reactor Power to Hot Standby
Revision 65
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous
Essential Raw Cooling Water - System 067, Maintenance
Rule (a)(1) Evaluation
October 2,
2019
Procedures
NEDP-8.0
Evaluation for Procurement of Materials, Items, and Services Revision 6
NPG-SPP-03.4
Maintenance Rule Performance Indicator Monitoring,
Trending and Reporting - 10CFR50.65
Revision 3
NPG-SPP-09.0.9
Conduct of Procurement Engineering
Revision 3
TI-4
Maintenance Rule Performance Indicator Monitoring,
Trending, and Reporting - 10CFR50.65
Revision 31
Miscellaneous
1-PE-2019/1-904-
0001C
Defense-in-depth protection list established for U1R23
10/14/2019
Procedures
1-PI-OPS-000-
20.2
Operator At The Controls Duty Station Checklists -Mode 5, 6
And Defueled, Attachment 2 SQN Defense in Depth
Assessment
Rev. 39
Miscellaneous
EWR 19-DEM-
074-119
Engineering evaluation to support IDO of RHR pump inlet
valve 1-FCV-74-3
10/28/2019
Engineering
Changes
PIC 23729
Revise DCN 22643A to reflect changes due to testing results Rev. A
Engineering
Changes
DCN 22703
Upgrade the Turbine Driven Auxiliary Feedwater Speed
Governor and Flow Controller
Revision 1
Procedures
0-SI-SXV-000-
206.0
Testing of Category A and B Valves after Work Activities,
Upon Release from a Hold Order, or when Transferred From
Other Documents
Rev. 7
0-SI-SXV-068-
201.0
Pressurizer PORV Operability Test
Rev. 2
1-SI-SXI-068-
201.0
Leakage Test of the Reactor Coolant Pressure Boundary
Rev. 15
1-SI-SXV-063-
205.0
Safety Injection Cold Leg Secondary Check Valve Integrity
Test
Rev. 12
Work Orders
119941800,
119941801,
20915536
Miscellaneous
Unit 1 Cycle 23 Outage Safety Plan
Revision 0
Procedures
0-GO-15
Containment Closure Control
Revision 48
0-PI-OPS-000-
Containment Inspection
Rev. 12
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
187.0
NPG-SPP-10.6
Infrequently Performed Test or Evolutions
Revision 1
Corrective Action
Documents
CR 1559305
Unable to maintain 12 psig on FSV-43-341 and FSV-43-317
10/22/2019
Engineering
Evaluations
EWR 19-EPG-
043-114
LLRT Pressure Drop across 200ft of 3/8 Polyfol Tubing
Rev. 1
Procedures
0-TI-SXX-068-
001.0
RCS Leakage Monitoring and Action Plans
Revision 5
Work Orders
119443660;
119619071;
20140791;
20261954
Perform 0-PI-SFT-360-001.0 Att 1 Placing Flex Diesel
Intermediate Pressure Pump In Service
08/27/2019
ALARA Plans
NPG-SPP-05.2.1
Opeational
ALARA Planning
and Controls
H8 Jack Shaft Replacement
Revision 1
Corrective Action
Documents
CR1480583;
CR1480512;
CR1522238;
Radiation
Surveys
4.04 Gamma ID
Sheet
Survey ID 102219013-2; Survey ID 102219011; Survey ID
2219014
10/23/2019;
10/22/2019;
10/22/2019
SQN-M-
20191019-14;
SQN-M-
20191022-6;
SQN-M-
20191021-16
10/19/2019
10:45;
10/22/2019
03:46;
10/21/2019
17:04
SQN-M-
20191023-13
CTM All R152 Remove & Replace H8 Drive Rod
10/23/2019
11:45
Corrective Action
CR 1559663
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Documents
71151
Corrective Action
Documents
1556565
Radiation
Surveys
SQN-M-
20191012-23
U1R23UC
10/12/2019
Radiation Work
Permits (RWPs)
RWP 19120002;
RWP 19140001
Lower Containment All Areas-High Radiation Area, High
Radiological Risk Activities ;
Upper Containment All Areas-Medium Radiological Risk
Activities
Revision 0;
Revision 0
Miscellaneous
Sequoyah Plant Health Committee Meeting Package
2/11/2019
Sequoyah 3rd Quarter 2019 - Site Trimester Performance
Assessment
June -
September
2019
SQN 2nd Quarter 2019 - Site Trimester Performance
Assessment
February -
May 2019
Corrective Action
Documents
Condition Reports (CRs) 1543714; 1544846;
Engineering
Changes
SQN-2-2018-067-
001
Temporary Swap of Function of 2-FCV-67-146 with 2-VLV-
67-551
Revision 0
Miscellaneous
SQN-0-19-099
PRA evaluation response for incorrect settings for 2-VLV-67-
551
09/18/2019
Operability
Evaluations
Past Operability Evaluation Documentation for CR 1544846
09/20/2019
Procedures
0-SO-67-1
Essential Raw Cooling Water
Revision 113
Local Operation of 2-VLV-67-551
Revision 13
NPG-SPP-06.9.3
Post-Modification Testing
Revision 11
NPG-SPP-09.5
Temporary Modifications, Temporary Configuration Changes Revision 16