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County Station Integrated Inspection Report 05000373/2021003; 05000374/2021003 and 07200070/2021001
	ML21309A760
Person / Time
Site:	LaSalle
Issue date:	11/08/2021
From:	Kenneth Riemer; NRC/RGN-III/DRP/B1
To:	Rhoades D; Exelon Generation Co, Exelon Nuclear
References
	IR 2021001, IR 2021003
	Download: ML21309A760 (33)
	v • d • e

SUBJECT:

LASALLE COUNTY STATION - INTEGRATED INSPECTION REPORT 05000373/2021003; 05000374/2021003 AND 07200070/2021001

Dear Mr. Rhoades:

On September 30, 2021, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at LaSalle County Station. On October 13, 2021, the NRC inspectors discussed the results of this inspection with Mr. P. Hansett, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.

Three findings of very low safety significance (Green) are documented in this report. Three 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.

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 III; the Director, Office of Enforcement; and the NRC Resident Inspector at LaSalle County Station.

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 III; and the NRC Resident Inspector at LaSalle County Station.

November 8, 2021 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, Kenneth R. Riemer, Chief Branch 1 Division of Reactor Projects Docket Nos. 05000373; 05000374; 07200070 License Nos. NPF-11 and NPF-18

Enclosure:

As stated

Inspection Report

Docket Numbers:

05000373; 05000374; 07200070

License Numbers:

NPF-11 and NPF-18

Report Numbers:

05000373/2021003; 05000374/2021003; 07200070/2021001

Enterprise Identifier:

I-2021-003-0092; I-2021-001-0161

Licensee:

Exelon Generation Company, LLC

Facility:

LaSalle County Station

Location:

Marseilles, IL

Inspection Dates:

July 01, 2021 to September 30, 2021

Inspectors:

G. Edwards, Health Physicist

R. Edwards, Senior Health Physicist

R. Elliott, Resident Inspector

O. Masnyk Bailey, Health Physicist

D. Sargis, Resident Inspector

W. Schaup, Senior Resident Inspector

R. Zuffa, Illinois Emergency Management Agency

Approved By:

Kenneth R. Riemer, Chief

Branch 1

Division of Reactor Projects

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at LaSalle County Station, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.

List of Findings and Violations

Failure to Survey to Ensure Occupational Doses Are as Low as Reasonably Achievable (ALARA)

Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000374/2021003-01 Open/Closed

[P.2] -

Evaluation 71124.02 A self-revealed finding of very low safety significance (i.e., Green), and an associated non-cited violation (NCV) of 10 CFR 20.1501 was identified by the inspectors involving the licensees failure to conduct radiological surveys to assure compliance with 10 CFR 20.1101(b) (ALARA). Specifically, the licensee failed to evaluate increasing radiological levels and potential radiological hazards (the accumulation of cobalt-based source term) from highly activated foreign material being released from degrading plant equipment (Unit 2 Reactor Recirculation Flow Control Valves) that was occurring from 2017 - 2021. The failure impeded the licensees ability to use procedures and engineering controls to achieve occupational doses ALARA, which is demonstrated by the greater than anticipated collective doses for several Unit 2 work activities.

Failure to Perform a Maintenance Risk Assessment that Included both Units for Replacing a Group 4 Isolation Relay Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000373,05000374/2021003-02 Open/Closed

[H.12] - Avoid Complacency 71152 The inspectors identified a finding of very low safety significance and an associated non-cited violation of Title 10 CFR 50.65(a)(4), "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," for the licensee's failure to assess and manage the increase in risk that may result from the proposed maintenance activities. Specifically, maintenance activities that replaced a Group 4 isolation relay on the shutdown unit were not assessed for the increased risk on the operating unit. This impacted the secondary containment of both units, when during a post-maintenance test (PMT), a Group 4 isolation occurred resulting in the loss of the reactor building ventilation system on the operating unit.

Diesel Generator Day Tank Manway Cover Leaks Due to Insufficient Torque Applied to Seating Bolts Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000373,05000374/2021003-03 Open/Closed None 71152 The inspectors documented a self-revealed Green finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control," for the licensee failing to review for suitability of application the process used to determine the torquing requirements to seat gaskets essential to the safety-related functions of the unit common diesel generator day tank manway. Specifically, when the licensee replaced the gasket for the manway cover on the unit common diesel generator day tank, the licensee failed to ensure the torque applied to the bolts to ensure proper seating stress for the gasket was sufficient to preclude a loss of bolt load due to gasket creep. The torque applied failed to ensure the gasket would be seated under all operating conditions and resulted in the manway cover leaking several weeks after the maintenance was completed, challenging the operability of the diesel generator.

Additional Tracking Items

None.

PLANT STATUS

Unit 1 began the inspection period at rated thermal power. On July 20, 2021, the unit performed an emergent down power to approximately 40 percent due to the loss of heater drain tank pump forward capability due to a failed controller. The failed controller was repaired, and the unit was returned to full-rated thermal power on July 21, 2021. On September 11, 2021, the unit was down powered to approximately 40 percent to add oil to the 'A' reactor recirculation pump oil reservoir, to perform reactor recirculation valve testing, and to perform a rod pattern adjustment. The unit was returned to full-rated thermal power on September 12, 2021, and the unit operated at or near rated thermal power for the remainder of the inspection period.

Unit 2 began the inspection period at rated thermal power. On September 18, 2021, the unit was down powered to approximately 83 percent to perform turbine control and stop valve testing, to perform scram time and channel distortion testing, and to perform a rod pattern adjustment. The unit was returned to full-rated thermal power on September 19, 2021, and the unit operated at or near rated thermal power 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 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), resident and regional inspectors were directed to begin telework and to remotely access licensee information using available technology. During this time, the resident inspectors performed periodic site visits each week, increasing the amount of time on site as local COVID-19 conditions permitted.

As part of their onsite activities, resident inspectors conducted plant status activities as described in IMC 2515, Appendix D; observed risk significant activities; and completed on site portions of IPs. In addition, resident and regional baseline inspections were evaluated to determine if all or a 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

71111.01 - Adverse Weather Protection

Impending Severe Weather Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the adequacy of the overall preparations to protect risk-significant systems for an extreme rain event of 6 to 8 inches of rainfall in 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months that occurred on July 13, 2021.

71111.04 - Equipment Alignment

Partial Walkdown Sample (IP Section 03.01) (2 Samples)

The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:

(1) Unit common diesel generator with the Unit 1 Division 2 diesel generator out for maintenance on August 30, 2021

(2) Unit 1 reactor core isolation cooling with high pressure core spray system out for maintenance on September 21, 2021

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 Samples)

The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:

(1) Fire Zone 4F3, auxiliary building, elevation 710'-0", ground floor (chemistry corridor)with unavailable fire risk important components on August 30, 2021

(2) Fire Zone 4E1, Unit 1 auxiliary building, elevation 731'-0", auxiliary equipment room with unavailable fire risk important components on September 21, 2021

(3) Fire Zone 3I1, Unit 2 reactor building, elevation 673'-4", general area (raceway) on September 28, 2021

(4) Fire Zone 5B13, Unit 2 auxiliary building, elevation 731'-0", balance of plant cable area on September 28, 2021

(5) Fire Zone 8B2, Unit 2 diesel building, elevation 710'-0", Division 2 standby diesel generator room and Fire Zone 8B4, Unit 2 diesel building, elevation 710'-0",

Division 2 diesel day tank room on September 28, 2021

Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the onsite fire brigade training and performance during an unannounced fire drill in conjunction with an emergency preparedness drill on July 22, 2021.

71111.06 - Flood Protection Measures

Inspection Activities - Internal Flooding (IP Section 03.01) (2 Samples)

The inspectors evaluated internal flooding mitigation protections in the:

(1) Unit 1 Unit common core standby cooling system pump room

(2) Unit 1 Division 2 core standby cooling system pump room

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 control room during a Unit 2 downpower to 83 percent rated thermal power to perform testing and make a rod pattern adjustment on September 18, 2021.

Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)

(1) The inspectors observed and evaluated out-of-the-box exam 3 on August 31, 2021.

71111.12 - Maintenance Effectiveness

Maintenance Effectiveness (IP Section 03.01) (3 Samples)

The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:

(1) Diesel generator day tanks

(2) Voltage regulators on the standby diesel generators

(3) Diesel generator lube oil coolers

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management Sample (IP Section 03.01) (4 Samples)

The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:

(1) Unit 1 Action Green online risk for Division 2 diesel generator maintenance on August 30, 2021

(2) Unit 1 Action Green online risk with the Unit 1 drywell de-inerted at power for a drywell entry to add oil to the 1A reactor recirculation pump on September 11, 2021

(3) Unit 2 Action Green online risk for Division 2 water leg pump maintenance on September 13, 2021

(4) Unit 1 Action Green online risk for Division 3 diesel generator maintenance (high pressure core spray system out of service) on September 20, 2021

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (6 Samples)

The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:

(1) Action Request 4437017, "Low Flow Discovered During LOS-DG-Q3"

(2) Action Request 4423492, "Unit 1 Division 1 VX (Switchgear Room Ventilation) Supply Fan Trip"

(3) Action Request 4444198, "Frequency of 1A Diesel Generator High During Operability Run"

(4) Action Request 4446308, "Replace Section of 2FC30AE-6" Pipe Based on UT Inspection"

(5) Engineering Change 353396, "Use of Roll Up Door on Turbine Deck, Elevation 786',

Between Turbine Building and Auxiliary Building on Column Row R North of Column R-11"

(6) Action Request 4448857, "1RH03CB-12 UT Void Upstream of 1E12-F053B"

71111.19 - Post-Maintenance Testing

Post-Maintenance Test Sample (IP Section 03.01) (5 Samples)

The inspectors evaluated the following post-maintenance test activities to verify system operability and functionality:

(1) Post-maintenance testing of 1VG003, standby gas treatment discharge isolation damper, on August 5, 2021 per Work Order 4800310

(2) Post-maintenance testing of the Unit 1 Division 2 diesel generator on September 1, 2021

(3) Post-maintenance testing of the Unit 1 Division 2 diesel generator lube oil cooler and fuel oil day tank on August 31 and September 1, 2021

(4) Post-maintenance testing of the Unit 2B residual heat removal pump after maintenance on September 14, 2021

(5) Post-maintenance testing of the Unit 1 Division 3 diesel generator on September 23, 2021

71111.22 - Surveillance Testing

The inspectors evaluated the following surveillance tests:

Surveillance Tests (other) (IP Section 03.01)

Main article: IP 71111.22

(1) LOS-VG-M1, Attachment 1A, Unit 1 standby gas treatment system operability and inservice test on August 5, 2021

(2) LTS-100-8, Attachment A, drywell personnel airlock door seal local leak rate test on September 12, 2021

(3) LOS-LP-Q1, Attachment 1A, Unit 1 low pressure core spray system operability and inservice test on September 8, 2021

(4) LOS-RH-Q1, Attachment 2B, Unit 2 residual heat removal system operability and inservice test on September 14, 2021

Inservice Testing (IP Section 03.01) (2 Samples)

(1) LOS-DG-Q3, Attachment A5, 1B diesel generator cooling water pump inservice test on July 26, 2021

(2) LOS-RH-Q1, Attachment 2A, Unit 2 2A residual heat removal system operability and inservice test on August 9, 2021

Inspection Review (IP Section 02.01-02.03) (1 Sample)

(1) The inspectors evaluated the following submitted Emergency Action Level and Emergency Plan changes.

Eval No. 20-75 Emergency Action Levels for LaSalle Station, EP-AA-1005, Addendum 3

Eval No. 20-78 Emergency Action Levels for LaSalle Station, EP-AA-1005, Addendum 3 This evaluation does not constitute NRC approval.

71114.06 - Drill Evaluation

Select Emergency Preparedness Drills and/or Training for Observation (IP Section 03.01) (3 Samples)

(1) LaSalle Station third quarter 2021 performance indicator (PI) drill, Emergency Response Organization (ERO) team B1, on July 14, 2021

(2) LaSalle Station third quarter 2021 PI drill, ERO team B2, on July 22, 2021

(3) LaSalle Station third quarter 2021 PI drill, ERO team D1, on August 4,

RADIATION SAFETY

71124.02 - Occupational ALARA Planning and Controls

Radiological Work Planning (IP Section 03.01) (3 Samples)

The inspectors evaluated the licensees radiological work planning.

(1) The inspectors evaluated items associated with Radiation Work Permit LA-02-21-00549 and associated as low as reasonably achievable (ALARA)documentation for L2R218 2B33-FO60A/B valve repairs/inspections during the spring 2021 outage.

(2) The inspectors evaluated items associated with Radiation Work Permit LA-02-21-00513 and associated ALARA documentation for L2R218 control rod drive activities during the spring 2021 outage.

(3) The inspectors evaluated items associated with Radiation Work Permit LA-02-21-00516 and associated ALARA documentation for L2R218 for drywell recirculation pump, seals, and motor activities during the spring 2021 outage.

Verification of Dose Estimates and Exposure Tracking Systems (IP Section 03.02) (3 Samples)

The inspectors evaluated dose estimates and exposure tracking.

(1) The inspectors evaluated items associated with Radiation Work Permit LA-02-21-00549 and associated ALARA documentation for L2R218 2B33-FO60A/B valve repairs/inspections during the spring 2021 outage.

(2) The inspectors evaluated items associated with Radiation Work Permit LA-02-21-00518 and associated ALARA documentation for L2R218 drywell inservice inspection (ISI) during the spring 2021 outage.

(3) The inspectors evaluated items associated with Radiation Work Permit LA-02-21-00901 and associated ALARA documentation for L2R218 reactor disassembly and reassembly activities during the spring 2021 outage.

71124.07 - Radiological Environmental Monitoring Program

Radiological Environmental Monitoring Program (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the implementation of the licensees radiological environmental monitoring program.

GPI Implementation (IP Section 03.03) (1 Sample)

(1) The inspectors evaluated the licensees implementation of the groundwater protection initiative program to identify incomplete or discontinued program elements.

71124.08 - Radioactive Solid Waste Processing & Radioactive Material Handling, Storage,

& Transportation

Radioactive Material Storage (IP Section 03.01)

Main article: IP 71124.08

(1) Inspectors evaluated the licensees performance in controlling, labelling and securing radioactive materials. This included the radioactive waste storage areas within the facility.

Radioactive Waste System Walkdown (IP Section 03.02) (1 Sample)

(1) Inspectors walked down accessible portions of the solid radioactive waste systems and evaluated system configuration and functionality.

Waste Characterization and Classification (IP Section 03.03) (2 Samples)

(1) The inspectors evaluated the licensees characterization and classification of radioactive waste. This included the resin waste streams at the facility.

(2) The inspectors evaluated the licensees characterization and classification of radioactive waste. This included the dry activated waste stream at the facility.

Shipment Preparation (IP Section 03.04) (1 Sample)

(1) The inspectors observed that a shipment containing radioactive material is prepared according to requirements. The shipment identification number was LW 21-018 for dewatered spent resin shipped on September 24, 2021.

Shipping Records (IP Section 03.05) (3 Samples)

The inspectors evaluated the following non-excepted radioactive material shipments through a record review:

(1) Shipment LW 21-014; Radioactive Material, Low Specific Activity (LSA-II), 7, UN3321; Dewatered Spent Resin; April 30, 2021

(2) Shipment LW 21-030; Radioactive Material, Low Specific Activity (LSA-II), 7, UN3321; Dry Activated Waste Sealand Container; April 22, 2021

(3) Shipment LW 21-018; Radioactive Material, Low Specific Activity (LSA-II), 7, UN3321; Dewatered Spent Resin; September 21,

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:

MS06: Emergency AC Power Systems (IP Section 02.05)===

Main article: IP 71151

(1) Unit 1 (July 1, 2020 through June 30, 2021)

(2) Unit 2 (July 1, 2020 through June 30, 2021)

MS07: High Pressure Injection Systems (IP Section 02.06) (2 Samples)

(1) Unit 1 (July 1, 2020 through June 30, 2021)

(2) Unit 2 (July 1, 2020 through June 30, 2021)

MS09: Residual Heat Removal Systems (IP Section 02.08) (2 Samples)

(1) Unit 1 (July 1, 2020 through June 30, 2021)

(2) Unit 2 (July 1, 2020 through June 30, 2021)

BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10) (2 Samples)

(1) Unit 1 (July 1, 2020 through August 31, 2021)

(2) Unit 2 (July 1, 2020 through August 31, 2021)

OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)

(1)

(July 1, 2020 through August 31, 2021)

PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample)

(1)

(July 1, 2020 through August 31, 2021)

71152 - Problem Identification and Resolution

Annual Follow-up of Selected Issues (IP Section 02.03) (2 Samples)

The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:

(1) Action Request 4407880, "Spurious Group 4 Signal Cause VG

[standby gas treatment]/VR [reactor building ventilation] Actuations"

(2) Action Request 4429239, "Unit Common Diesel Generator Day Tank Leaking Fuel Oil"

OTHER ACTIVITIES

- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL 60855 - Operation of an ISFSI The inspectors evaluated the licensees independent spent fuel storage installation (ISFSI)cask loading activities August 9 through August 26, 2021. Specifically, the inspectors observed the following activities during the loading of multi-purpose canister (MPC) Nos.

603 and 665:

Heavy lift of the loaded transfer cask (HI-TRAC) from the spent fuel pool to the cask processing area

Closure welding and non-destructive evaluations

Troubleshooting of a remote valve operating assembly

Canister processing including forced helium dehydration and backfill with helium

Stack-up and download of the MPC from the HI-TRAC into the storage cask (HI-STORM)

Transport of a loaded HI-STORM out to the ISFSI pad

Radiological field surveys The inspectors performed walkdowns of the ISFSI pad, including walkdowns of the ISFSI haul path.

The inspectors evaluated the following:

Spent fuel selected for loading into dry cask storage during this loading campaign

Selected corrective action program documents

Selected 72.48 screenings and evaluations

INSPECTION RESULTS

Failure to Survey to Ensure Occupational Doses Are as Low as Reasonably Achievable (ALARA)

Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000374/2021003-01 Open/Closed

[P.2] -

Evaluation 71124.02 A self-revealed finding of very low safety significance (i.e., Green) and an associated non-cited violation (NCV) of 10 CFR 20.1501 was identified by the inspectors involving the licensees failure to conduct radiological surveys to assure compliance with 10 CFR 20.1101(b) (ALARA). Specifically, the licensee failed to evaluate increasing radiological levels and potential radiological hazards (the accumulation of cobalt-based source term) from highly activated foreign material being released from degrading plant equipment (Unit 2 Reactor Recirculation Flow Control Valves) that was occurring from 2017-2021. The failure impeded the licensees ability to use procedures and engineering controls to achieve occupational doses ALARA, which is demonstrated by the greater than anticipated collective doses for several Unit 2 work activities.

Description:

During the LaSalle Unit 2 spring 2017 outage, the licensee engaged in a collective radiation exposure reduction effort which included chemical decontamination of several reactor systems, suppression pool and guide tube vacuuming, core vacuuming (with a full core offload), and high efficiency ultra-sonic fuel cleaning on several bundles of fuel that were loaded into the core. During this effort, the licensee also identified previously undetected highly activated foreign material in several areas of the reactor, including the bottom head drain, fuel debris screens and the annulus floor. The foreign material caused elevated dose rates in different areas of the reactor, including but not limited to, the under-vessel area, a hotspot on the double block drain valves (which had on contact dose rates of 44 Rem/hr and 2.6 Rem/hr at 1 foot) and elevated dose rates during the initial drywell head removal.

The licensee collected portions of the foreign material and sent it for analysis to determine the chemical composition of the material and fracture patterns that were present in the material in fall of 2017. The results revealed that multiple pieces of the material were composed of wear resistant cobalt-based alloys and that there was a high probability that the material originated from same component. When the licensee received the results, station engineering performed an assessment of all the flow paths to the reactor cavity which could result in the transmission of foreign material migrating to the lower plenum area of the reactor. The analysis concluded that there was a high probability that the foreign material originated from the reactor recirculation system. The licensee engineering staff then determined that the material originated from the reactor recirculation (RR) ball valves (BVs) or the reactor recirculation flow control valves (FCVs) given that these valves have components of similar geometry and chemical composition. Additionally, the licensee's analysis concluded that the RR FCVs contained several Stellite 6 (composed of cobalt-based alloys) bearings and spacer rings and that pieces of the foreign material closely matched spacers that were in storage. During the summer of 2018, the licensees engineering staff made a recommendation to inspect the RR FCVs during the next outage (spring 2019) to determine if the valves were the source of the foreign material. The recommendation by the engineering staff focused on the elevated risk of future debris related fuel failures and did not address the effects that the foreign material would have to the units source term and collective radiation exposure. The recommendation to inspect the RR FCVs during the spring 2019 outage was not accepted by the site management.

During the spring 2019 outage, boiling water reactor radiation level assessment and control (BRAC) dose rates had risen from 35 mRem/hr (post-Chemistry Decontamination) to 250 mRem/hr, which confirmed that increased levels of cobalt were plating out and accumulating inside of the reactor systems. The licensee performed fuel cleaning and identified that approximately 29 percent of the cleaned fuel bundles contained highly activated foreign material. Foreign material was also identified in several jet pumps (including the jet pump risers), reactor water clean-up valves, the bottom head drain and the annulus floor. The foreign material was unable to be evaluated to determine the source of the material due to dose rates exceeding 4000 Rem/hr on contact.

Through 2019 - 2021, the licensee was able to identify through chemical analysis that cobalt introduction into the system continued to increase. Despite these trends, the licensee failed to evaluate how the increased concentrations would affect the units source term and the collective radiation exposure and how the continued increase of cobalt in the reactor would affect dose rates in the spring 2021 outage. When the spring 2021 outage commenced, BRAC dose rates were approximately 455 mRem/hr, which is an increase of 205 mRem/hr from the previous outage (spring 2019), and dose rates in the drywell had also increased from 131 percent - 190 percent. Although this information showed a significant increase in dose rates, the licensee decided to not adjust any of the ALARA work scopes, dose reduction strategies, or dose estimations until work on components within the drywell had begun.

The outage job with the largest dose contributing to the spring 2021 outage was the inspection/repair of the RR FCVs. When the licensee began inspections on the RR FCVs, the extensive damage to the valves confirmed that the valves were the source of the highly activated foreign material that had been identified throughout the reactor. The licensee determined that that valves would need extensive repairs due to damage that they had endured over several operating cycles. Although the licensee had determined that the valves need to be inspected in previous years, the licensee did not recognize and/or consider that the valves would need repairs and, therefore, did not incorporate adequate ALARA plans and/or controls in relation to repairs of the RR RCVs. The licensee only evaluated disassembly, decontamination, inspection, and reassembly of the RR FCVs. The failure to incorporate repair plans into the initial ALARA work planning caused unplanned and unanticipated collective dose from work conducted on the valves. Specifically, the licensee failed to incorporate processes such as using mock-ups to simulate the valve repairs for the repairs associated with these valves. The initial dose estimate for RR FCVs work was 28 Rem and the work was completed for 306 Rem. Additionally, several other radiologically controlled jobs during the spring 2021 were completed with doses that exceeded 5 Rem and greater than 50 percent of the initial work estimate. The spring 2021 outage was completed for 664 Rem on an initial estimate of 235 Rem.

As described above, an unusual amount of cobalt, and consequently the amount available to become activated, was evident as early as 2017. In 2018, members of the licensee staff suspected the source of material was the RR FCVs. However, they failed to act, such as inspecting the valves during the 2019 refueling outage, even though routine on-line chemistry analysis indicated that amount of activated cobalt was increasing the system. This on-line data might have predicted the dramatically higher dose rates observed in the 2019 refueling outage, although it seemed unexpected when the conditions were observed. After the 2019 outage, the on-line chemistry data continued to show an increase in activated cobalt in the system and the licensee did not use this data during the planning for the 2021 refueling outage work activities. For these reasons, the inspectors determined that the failure to evaluate increasing radiological levels and potential radiological hazards was self-revealed.

Corrective Actions: The licensee was considering revising ER-AA-2006 Lost Parts Evaluation Process to evaluate foreign material for source term impacts to the facility and to include a process that will connect Radiation Protection, Design Engineering, and Chemistry to evaluate these lost parts for source term impact using RP-AA-551 Cobalt Reduction Program.

Additionally, the licensee planned to remove and mitigate the amount of highly activated foreign material that remains in the Unit 2 reactor. These corrective actions include and are not limited to the identification of components composed of Stellite and analyzing their impact on the units source term and plans for future removal, replacement of the bottom head drain, future use of high-efficiency ultrasonic fuel cleaning and future chemical decontamination efforts on the unit.

Corrective Action References: Issue Report 4417704; "Elevated Collective Radiation Exposure on Unit 2 Resulting in a Total Outage Dose of 664.718 Rem on an Original L2R18 Business Plan Dose Goal of 235 Rem"; dated June 17, 2021

Performance Assessment:

Performance Deficiency: The licensee did not make surveys (as required by 10 CFR 20.1501) to assure compliance with 10 CFR 20.1101

(b) ALARA. Specifically, the licensee failed to evaluate increasing radiological levels and potential radiological hazards (the accumulation of cobalt based source term) from highly activated foreign material being released from degrading plant equipment (Unit 2 RR FCVs) that was occurring from 2017-2021, which impeded the licensees ability to use procedures and engineering controls to achieve occupational doses ALARA.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Program & Process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the licensee did not effectively implement procedures or engineering controls to achieve doses that are ALARA as indicated by actual collective dose exceeding 5 person-Rem AND exceeding the planned (or adequately re-planned), intended dose by more than 50 percent.

Significance: The inspectors assessed the significance of the finding using Appendix C, Occupational Radiation Safety SDP. The inspectors assessed the significance of the finding using Appendix C, Occupational Radiation Safety SDP. The inspectors determined that the finding was of very low safety significance (i.e., Green) because: The finding involved ALARA planning and work controls but the licensee's current 3-year (2018 - 2020) rolling average collective dose is less than or equal to 240 person-Rem/unit for a boiling water reactor (BWR) so the significance of the inspection finding is Green. Additionally,

(1) there was no overexposure,

(2) there was no substantial potential for an overexposure, and

(3) the ability to assess dose was not compromised.

Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee had evidence that supported the inspection and repair of the RR FCV during the spring 2019 outage, station management made the decision to delay the inspections while not evaluating or fully understanding the consequences to material conditions, plant source term and collective radiation exposure from identified highly activated foreign material.

Enforcement:

Violation: Title 10 CFR 20.1501 requires that each licensee make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in Part 20 and that are reasonable under the circumstances to evaluate the extent of radiation levels, concentrations or quantities of radioactive materials, and the potential radiological hazards that could be present.

Title 10 CFR 20.1101(b) requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are ALARA.

Title 10 CFR 20.1003 Definitions, states, as used in this part:

Survey means an evaluation of the radiological conditions and potential hazards incident to the production, use, transfer, release, disposal, or presence of radioactive material or other sources of radiation. When appropriate, such an evaluation includes a physical survey of the location of radioactive material and measurements or calculations of levels of radiation, or concentrations or quantities of radioactive material present.

ALARA (acronym for "as low as is reasonably achievable") means making every reasonable effort to maintain exposures to radiation as far below the dose limits in this part as is practical consistent with the purpose for which the licensed activity is undertaken, taking into account the state of technology, the economics of improvements in relation to state of technology, the economics of improvements in relation to benefits to the public health and safety, and other societal and socioeconomic considerations, and in relation to utilization of nuclear energy and licensed materials in the public interest.

Contrary to the above, during the time period of 2017 to 2021, the licensee did not make surveys to assure compliance with 10 CFR 20.1101(b), which requires the use of procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses that are ALARA. Specifically, the licensee failed to evaluate increasing radiological levels and potential radiological hazards (the accumulation of radiological source term) from highly activated foreign material being released from degrading plant equipment (Unit 2 RR FCVs) that was occurring from 2017 to 2021. The failure to evaluate the radiological hazards impeded the licensees ability to use procedures and engineering controls to achieve occupational doses ALARA, which is demonstrated by the greater than anticipated occupational doses for several Unit 2 work activities.

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

Failure to Perform a Maintenance Risk Assessment that Included both Units for Replacing a Group 4 Isolation Relay Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000373,05000374/2021003-02 Open/Closed

[H.12] - Avoid Complacency 71152 The inspectors identified a finding of very low safety significance and an associated non-cited violation of Title 10 CFR 50.65(a)(4), "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," for the licensee's failure to assess and manage the increase in risk that may result from the proposed maintenance activities. Specifically, maintenance activities that replaced a Group 4 isolation relay on the shutdown unit were not assessed for the increased risk on the operating unit. This impacted the secondary containment of both units, when during a post-maintenance test (PMT), a Group 4 isolation occurred resulting in the loss of the reactor building ventilation system on the operating unit.

Description:

On March 10, 2021, at 0815 hours0.00943 days 0.226 hours 0.00135 weeks 3.101075e-4 months , the Unit 1 and Unit 2 control rooms received multiple alarms indicating that a Group 4 isolation had occurred and that both standby gas treatment systems had started. Unit 2 was shut down with fuel moves occurring and Unit 1 was operating at full-rated thermal power. Fuel moves were immediately halted on Unit 2, and Unit 1 operators entered off-normal procedures for a loss of reactor building ventilation, a system that supports secondary containment. At the time of the event, Unit 2 reactor building ventilation was isolated for outage activities and the electrical maintenance department was installing test equipment on Unit 2 to test a relay (C71-K4B1) associated with the Group 4 isolation logic. While installing the test equipment, a technician shorted a test connection across two terminals resulting in a blown fuse and a Group 4 isolation on both units. The Group 4 isolation secured the reactor ventilation on the operating unit as designed, resulting in a momentary loss of secondary containment until both trains of the standby gas treatment system initiated, restoring secondary containment.

On February 28, 2021, the licensee determined that the initial PMT selected for the relay maintenance would not need to be performed and would save 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months of critical path for the outage and that a special PMT would need to be developed to test the replaced relay. On March 8, 2021, during the first performance of the special PMT, the technicians identified that there were several fuses that needed further identification in the work package. After revising the work package to identify the fuses, on March 9, 2021, a second attempt was made to perform the PMT. During this attempt it was identified that some of the jumpers in the work package could not be used because they would defeat functions required to be tested to ensure the relay replaced would function. After revising the work package to ensure the appropriate jumpers were installed, on March 10, 2021, a third attempt was made at the PMT. During this test the technicians shorted the terminals, blowing a fuse, causing the Group 4 isolation signal and loss of the reactor building ventilation system on the operating unit.

The inspectors requested all work orders and associated maintenance paperwork and a completed copy of the corrective action program evaluation (CAPE) performed. The inspectors determined the following:

The short occurred due to the technicians using skill-of-the-craft techniques to mitigate possible contact with terminals while installing jumpers. As a robust barrier, the technicians could have decided to use electrical tape to cover the adjacent terminals where the jumper was to be landed, mitigating the possibility of a short. The technicians in the field and the supervisor overseeing the task did not identify this human performance tool during the pre-job brief or during the 2-minute drill. The licensee stated in the CAPE that the risk versus benefit was not properly assessed as part of their robust barriers program.

The licensee performed a pre-job brief each day; however, different technicians were on the job different days. The level of detail at each brief was not consistent and did not cover issues identified during the previous attempts to perform the procedure.

The CAPE also identified that the planners and engineers that developed the PMT did not implement proper technical rigor when, after thoroughly reviewing the prints, it was discovered that banana jack points existed further down the circuit that performed the same function as the jumper that was being installed. These jacks could have been used to ease jumper installation and provide high reliability that the jumpers would remain in place.

The inspectors asked the licensee if the maintenance activity had been screened in accordance with station procedure WC-AA-101-1006, "On-line Risk Management and Assessment," for the operating unit and if OP-AA-107F-01, "Risk Screens/Mitigation Plan," had been completed with a risk mitigation plan for the maintenance activity.

These procedures are used by the station for maintenance activities covered under 10 CFR 50.65(a)(4) which states, in part, "Before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities." The inspectors were told that the station had not assessed the risk in accordance with WC-AA-101-1006 nor was a risk mitigation plan implemented for the operating unit.

Corrective Actions: The licensee documented the event in Action Request 4407880 and performed a CAPE of the event. Immediate corrective actions included operations restoring reactor building ventilation on the operating unit and changing the work package to utilize the banana jacks for the PMT performance. The PMT was completed successfully. The licensee documented findings and corrective actions in the CAPE.

Corrective Action References: Action Request 4407880

Performance Assessment:

Performance Deficiency: The inspectors determined that the failure to assess maintenance risk on the operational unit for the replacement of the Group 4 isolation relay on the shutdown unit as required by Title 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the loss of the reactor building ventilation system during the Group 4 isolation challenged secondary containment for both units and required the standby gas treatment systems to start to maintain secondary containment.

Significance: The inspectors assessed the significance of the finding using Appendix K, Maintenance Risk Assessment and Risk Management SDP. The inspectors discussed the performance deficiency with a senior reactor analyst and determined the following. The loss of the reactor building ventilation system impacts the ability of the secondary containment to perform its function. Licensee station document LS-PRA-015, "LaSalle County Generation Station Probabilistic Risk Assessment, Level 2/LERF Notebook," Revision 7, describes the Exelon Level 2 probabilistic risk assessment (PRA) methodology for LaSalle County Generating Station (LCGS) deriving the radionuclide release categories and frequencies that characterize the severe accident spectrum. The notebook describes what the secondary containment function is and how it was evaluated as part of the large early release frequency. The appendix discusses determining the risk deficit for the event and then reviewing the net risk deficit to determine the significance of the finding. The Level 2 PRA provides the following information about how the secondary containment is evaluated for the large early release frequency.

For the secondary containment to retain a significant quantity of fission products, one of two conditions must occur.

First, in many cases what might be loosely referred to as "active" decontamination measures should be available. This would include scrubbing due to the passage of fission products through deep water pools, decontamination by ventilation system filters, or scrubbing due to wide-coverage fire sprays. If such measures are functional, they would generally overwhelm the natural settling processes and result in relatively small environmental releases of all fission products except for noble gases. However, a few qualifications to this statement must be offered. First, ventilation filters are not usually designed for the large aerosol loadings that would be seen in a severe accident and consequently may tear, overheat, or clog. Second, fire sprays may not cover the area of all the affected secondary containment regions. Finally, while aerosol behavior is relatively well understood, there are significant uncertainties associated with the effectiveness of scrubbing fission product vapors in water pools; these might impact the release when the source of fission products is at a very high temperature.

If no such active measures are at work, a natural settling process must be relied upon. For this to be effective, the fission products may require a relatively long residence time in the secondary containment before they can be swept into the environment. This in turn requires that the ventilation systems be secured, that the flowrate from the primary system or containment be relatively small, and that vigorous natural circulation be avoided between the secondary containment and the environment. The last of these requirements is often the most difficult to confirm. Vigorous natural circulation between the secondary containment and the environment can be set up if one large hole is opened (leading to large counter-current flows through the one opening), or if two holes are opened, one low in the building and one higher up. This latter configuration gives rise to a "chimney-like" flow pattern.

Consequently, given the possible different removal mechanisms and the associated effectiveness (or lack of effectiveness) of each, the reactor building could be examined in the PRA model. However, for this evaluation, the reactor building is given very little weight as a mitigation feature. This may be conservative.

Based upon the above, it was determined that for the event no change in risk deficit occurred and therefore screens as Green in accordance with Flowchart 1, Assessment of Risk Deficit of IMC 0609 Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process.

Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. Specifically, the operations department was complacent when they failed to validate that integrated risk had been properly assessed and mitigated in accordance with station procedures and relied on the shutdown management plan for both units.

Enforcement:

Violation: Title 10 CFR 50.65(a)(4) states, in part, "Before performing maintenance activities (including but not limited to surveillance, PMT, and corrective and preventive maintenance),the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities."

Contrary to the above on March 10, 2021, before performing maintenance activities, the licensee failed to assess and manage the increase in risk for activities that resulted from maintenance activities. Specifically, maintenance activities that replaced a Group 4 isolation relay on the shutdown unit were not assessed for the increased risk on the operating unit. This impacted the secondary containment of both units, when during a PMT, a Group 4 isolation occurred resulting in the loss of the reactor building ventilation system.

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

Diesel Generator Day Tank Manway Cover Leaks Due to Insufficient Torque Applied to Seating Bolts Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000373,05000374/2021003-03 Open/Closed None 71152 The inspectors documented a self-revealed Green finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control," for the licensee failing to review for suitability of application the process used to determine the torquing requirements to seat gaskets essential to the safety-related functions of the unit common diesel generator day tank manway. Specifically, when the licensee replaced the gasket for the manway cover on the unit common diesel generator day tank, the licensee failed to ensure the torque applied to the bolts to ensure proper seating stress for the gasket was sufficient to preclude a loss of bolt load due to gasket creep. The torque applied failed to ensure the gasket would be seated under all operating conditions and resulted in the manway cover leaking several weeks after the maintenance was completed, challenging the operability of the diesel generator.

Description:

On June 14, 2021, operations was contacted by a security officer that the unit common diesel generator day tank was leaking fuel oil. Operations dispatched operators to the day tank room. The operators identified the source of the leak as the manway gasket on the day tank and quantified the leak as approximately 1 gpm. The operators initiated action to contain the leak, and maintenance personnel were able to tighten the manway cover to reduce leakage to approximately 1/4 gpm. The unit common diesel generator was taken out of service and the rubber sheet gasket that leaked was replaced with a gore power grade sheet gasket. The diesel generator was returned to service and the licensee documented the event as Action Request 4429239.

The inspectors reviewed the work order and corresponding engineering documents and determined the following:

The work instructions that replaced the rubber sheet gasket on May 10, 2021, included instructions to torque the manway access cover to 14 ft-lb. The torque valve had been previously determined by station calculation L-002533 and incorporated into the work instructions. After the event, engineering reviewed this torque valve at the request of the maintenance department to verify that it was sufficient to seat the gasket properly.

Engineering determined that 14 ft-lb resulted in a seating stress of 644 psi, which was slightly above the minimum required seating stress of 600 psi. Although these values suggest that the gasket was properly seated, there was little margin for loss of bolt load due to gasket creep. A loss of only 1 ft-lb drops the gasket seating stress to 599 ft-lb which is below the minimum required seating stress per the vendor specifications.

EPRI Technical Report 3002008061, "Assembling Gasketed, Flanged Bolted Joints,"

Revision 1, states that rubber sheet gaskets are susceptible to "high degree of bolt relaxation." Since sufficient gasket stress is necessary to prevent leakage, loss of bolt preload due to gasket creep decreases the gasket seating stresses, thus increasing the probability of joint leakage. The torque valve provided in the work instructions did not take these factors into account, resulting in the joint leaking and challenging the operability of the diesel generator.

Corrective Actions: The licensee determined that a more suitable material could be used for the gasket and provided an engineering change to maintenance that would ensure that sufficient torque would be applied to seat the new material preventing leakage. Additionally, the licensee checked the remaining diesel generator day tanks as part of the extent of condition and has work requests generated to replace the remaining gaskets during scheduled maintenance activities.

Corrective Action References: Action Requests 4429239 and 4446082

Performance Assessment:

Performance Deficiency: The inspectors determined that the licensee's failure to review for suitability of application of the process used to determine the torquing requirements to seat gaskets essential to the safety-related functions of the unit common diesel generator day tank manway in accordance with 10 CFR Part 50, Appendix B, Criterion III, Design Control, was a performance deficiency.

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. Specifically, not providing torque requirements to ensure the day tank manway cover would not leak, challenged the diesel's reliability under all operating conditions.

Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors screened the issue against the Mitigating System questions and determined the finding was a very low safety significance (Green) since the finding was not a deficiency affecting the design or qualification of a mitigating SSC, did not represent a loss of system and/or function, and did not represent an actual loss of function.

Cross-Cutting Aspect: None

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components.

Contrary to the above, on May 10, 2021, the licensee failed to review for suitability of application of the process used to determine the torquing requirements to seat gaskets essential to the safety-related functions of the unit common diesel generator day tank manway. Specifically, when the licensee replaced the gasket for the manway cover on the unit common diesel generator day tank, the licensee failed to ensure the torque applied to the bolts to ensure proper seating stress for the gasket was sufficient to preclude a loss of bolt load due to gasket creep. The torque applied failed to ensure the gasket would be seated under all operating conditions and resulted in the manway cover leaking several weeks after the maintenance was completed, challenging the operability of the diesel generator.

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

Observation: Spurious Group 4 Signal Causes VG/VR Actuations 71152 The inspectors reviewed Action Request 4407880, "Spurious Group 4 Signal Causes VG/VR Actuations," as a sample for annual selected issue(s) for follow-up focusing on the following performance attributes of IP 71152:

Complete, accurate, and timely documentation in the corrective action program

Evaluation and timely disposition of operability and reportability issues

Consideration of the extent of condition and cause, generic implications, common cause, and previous occurrences

Evaluation and timely disposition of operability and reportability issues

Classification and prioritization of the resolution of the problem commensurate with safety significance

Identification of corrective actions, which were appropriately focused to correct the problem

Completion of corrective actions in a timely manner commensurate with the safety significance of the issue

Identification of negative trends associated with human or equipment performance that can potentially impact nuclear safety

Operating experience is adequately evaluated for applicability and applicable lesson learned are communicated to appropriate organizations and implemented The inspectors determined that the licensee had appropriately followed station procedures and the station's corrective action program to ensure all elements inspected were adequately addressed. As part of the review, the inspectors determined that the licensee failed to assess and mitigate risk with the work to replace a relay related with the Group 4 isolation on the operating unit. This issue was dispositioned as an NCV of 10 CFR 50.65(a)(4) in this report.

Observation: Unit Common Diesel Generator Day Tank Leaking Fuel Oil 71152 The inspectors reviewed Action Request 4429239, "0 Diesel Generator Day Tank Leaking Fuel Oil," as a sample for annual selected issue(s) for follow-up focusing on the following performance attributes of IP 71152: