Text

County Station - Integrated Inspection Report 05000373/2020001 and 05000374/2020001
	ML20133J813
Person / Time
Site:	LaSalle
Issue date:	05/11/2020
From:	Kenneth Riemer; Reactor Projects Branch 1
To:	Bryan Hanson; Exelon Generation Co, Exelon Nuclear
References
	IR 2020001
Preceding documents:	ML19337E206 (3 December 2019);
	Download: ML20133J813 (31)
	v • d • e

May 11, 2020

SUBJECT:

LASALLE COUNTY STATION - INTEGRATED INSPECTION REPORT 05000373/2020001 AND 05000374/2020001

Dear Mr. Hanson:

On March 31, 2020, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at LaSalle County Station. On April 8, 2020, the NRC inspectors discussed the results of this inspection with Mr. P. Hansett, Plant Manager, and other members of your staff. The results of this inspection are documented in the enclosed report.

One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. One Severity Level IV violation without an associated finding is documented in this report. 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. 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/

Kenneth R. Riemer, Chief Branch 1 Division of Reactor Projects

Docket Nos. 05000373 and 05000374 License Nos. NPF-11 and NPF-18

Enclosure:

As stated

Inspection Report

Docket Numbers:

05000373 and 05000374

License Numbers:

NPF-11 and NPF-18

Report Numbers:

05000373/2020001 and 05000374/2020001

Enterprise Identifier: I-2020-001-0048

Licensee:

Exelon Generation Company, LLC

Facility:

LaSalle County Station

Location:

Marseilles, IL

Inspection Dates:

January 01, 2020 to March 31, 2020

Inspectors:

J. Benjamin, Senior Reactor Inspector

G. Edwards, Health Physicist

N. Feliz-Adorno, Senior Reactor Inspector

J. Havertape, Resident Inspector

C. Hunt, Resident Inspector

J. Neurauter, Senior Reactor Inspector

E. Sanchez Santiago, Senior Reactor Inspector

W. Schaup, Senior Resident Inspector

C. St. Peters, Reactor Engineer

L. Torres, 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

Primary Containment Isolation Valve was not Properly Qualified for Submergence Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000373/2020001-01 Open/Closed None (NPP)71152 The inspectors identified a finding of very low safety significance and an associated non-cited violation (NCV) of 10 CFR 50.49, "Environmental qualification of electric equipment important to safety at nuclear power plants," for the failure of the program for qualifying the electric equipment important to safety to prevent submergence. Specifically, during a steam leak, water filled the motor, limit switch compartment, flex conduit, and junction box for a primary containment isolation valve, resulting in the valve becoming inoperable due to submergence.

Condition Prohibited by Technical Specifications Due to Turbine Stop Valve Limit Switch Failure Cornerstone Significance Cross-Cutting Aspect Report Section Not Applicable NCV 05000373/2020001-02 Open/Closed Not Applicable 71153 The inspectors reviewed a self-revealed severity level IV (SLIV) NCV of 10 CFR 50.36(c)(2)(i),

Limiting Conditions for Operation [LCO], for failing to meet/follow the required actions for LCO 3.3.1.1, Reactor Protection System (RPS) Instrumentation, and 3.3.4.1, End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation. Specifically, operators performed surveillance testing per station procedure LOS-RP-Q2 on May 20, 2018, and subsequently on September 9, 2018, during which turbine stop valve limit switch #3 was found inoperable. The licensee later determined after a failure analysis of the limit switch that during the May 20, 2018, surveillance that turbine stop valve #3 limit switch, 1C71-N006C, was actually failed rendering the associated RPS and EOC-RPT channels inoperable on May 20, 2018, even though corrective actions were taken and the limit switch was declared operable at that time.

Additional Tracking Items

Type Issue Number Title Report Section Status URI 05000373,05000374/20 19012-01 Main Steam Isolation Valve Fast Stoke Time Test 71111.21M Closed LER 05000373/2019-001-00 LER 2019-001-00 for LaSalle County Station, Unit 1, Turbine Stop Valve Limit Switch Failure due to Lubricant Degradation.

71153 Closed

PLANT STATUS

Unit 1 began the inspection period in coast down to refueling outage L1R18. On January 29, 2020, the unit was down powered to 60 percent to perform planned troubleshooting of the off-gas system. The down power was required to reduce radiation levels in areas throughout the plant (e.g. heater bays, main condenser, steam tunnel, etc.) to lower dose rates for personnel entries into the areas as part of the troubleshooting. The unit was returned to full power the following day. On February 9, 2020, Unit 1 was shut down to commence refueling outage L1R18. On February 26, reactor plant startup commenced, and the unit reached rated thermal power. The unit remained at or near rated thermal power for the remainder of the inspection period.

Unit 2 began the inspection period at rated thermal power. On March 14, 2020, the unit was down powered to approximately 72 percent to perform turbine stop valve, feedwater valve main steam isolation valve testing, scram time testing, rod sequence exchange, and rod pattern adjustment. The unit was returned to rated thermal power the following day. The unit remained 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/readingrm/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. From January 1 - March 19, 2020, 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.

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 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 and during that time conducted plant status activities as described in IMC 2515, Appendix D; and observed risk significant activities when warranted. In addition, resident and regional baseline inspections were evaluated to determine if all or portion of the objectives and requirements stated in the IP could be performed remotely. If the inspections could be performed remotely, they were conducted per the applicable IP. In the cases where it was determined the objectives and requirements could not be performed remotely, management elected to postpone and reschedule the inspection to a later date.

REACTOR SAFETY

71111.04 - 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 Division 2 emergency diesel generator during Yellow shutdown safety risk window on February 18, 2020

(2) Unit 1 standby liquid control system after system restoration from maintenance on February 24, 2020

(3) Unit 1 standby gas treatment system after surveillance testing (single train) on February 25, 2020

Complete Walkdown Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated system configurations during a complete walkdown of the Unit 2 reactor core isolation cooling system on January 24, 2020

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 4E2, auxiliary building, elevation 731', Unit 2 auxiliary electrical equipment room on January 3, 2020

(2) Fire Zone 4F2, auxiliary building, elevation 710', Unit 2 Division 1 essential switchgear room on January 3, 2020

(3) Fire Zone 4E4, auxiliary building, elevation 731', Unit 2 Division 2 essential switchgear room on January 3, 2020

(4) Fire Zone 4B, auxiliary building, elevation 786', lower ventilation equipment floor on January 16, 2020

(5) Fire Zone 4A, auxiliary building, elevation 815', upper ventilation equipment floor on January 21, 2020

71111.07A - Heat Sink Performance

Annual Review (IP Section 03.01) (1 Sample)

The inspectors evaluated readiness and performance of:

(1) Unit 1 'A' residual heat removal heat exchanger

71111.08G - Inservice Inspection Activities (BWR)

BWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding

Activities (IP Section 03.01)

Main article: IP 71111.08G

(1) The inspectors verified the reactor coolant system boundary, reactor vessel internals, risk-significant piping system boundaries, and containment boundary were appropriately monitored for degradation and repairs and replacements were appropriately fabricated, examined and accepted by reviewing the following activities from February 10, 2020 to April 6, 2020

03.01.a - Nondestructive Examination and Welding Activities.

1. Volumetric, Ultrasonic Examination (UT) of Feedwater System piping weld,

American Society of Mechanical Engineers (ASME) Category R-A, Component 1FW-1001-10 located on Line 1FW02FA-24"

2. Volumetric, UT of Feedwater System piping weld, ASME Category R-A,

Component 1FW-1001-13 located on Line 1FW02FA-24"

3. Volumetric, UT of Feedwater System piping weld, ASME Category R-A,

Component 1FW-1001-16 located on Line 1FW02FA-24"

4. Volumetric, UT of Feedwater System piping weld, ASME Category R-A,

Component 1FW-1002-10 located on Line 1FW02FB-24"

5. Volumetric, UT of Feedwater System piping weld, ASME Category R-A,

Component 1FW-1002-16, located on Line 1FW02FB-24"

6. Surface, Magnetic Particle Examination (MT) of Reactor Pressure Vessel

support welded attachment, ASME Category B-K, Component IVS-1-2-3 located on vessel skirt

7. Visual Examination (VT-3) of Reactor Recirculation System pump restraint,

ASME Category F-A, snubber Component RR00-1013S

8. Visual Examination (VT-3) of Reactor Recirculation System pump restraint,

ASME Category F-A, snubber Component RR00-1015S

9. Visual Examination (VT-3) of Reactor Pressure Vessel support, ASME

Category F-A, vessel skirt, Component 1B13-D003 10. Feedwater System, Line 1FW11A-4 Permanent Repair Downstream of Valve 1G33-F040, Weld Numbers 1A, 2, 3, 4, and 5 (Work Order 4943447-01)

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 the Unit 1 reactor shutdown on February 10, 2020, and the Unit 1 startup on February 27, 2020

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

(1) The inspectors observed and evaluated an out of the box examination (OBE) in the simulator on March 31, 2020

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management Sample (IP Section 03.01) (1 Sample)

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 Yellow shutdown safety risk window due to unit common diesel generator voltage regulator replacement and Division 1 alternating current bus work on February 18, 2020

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)

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

(1)insulation removal from Unit 1 reactor core isolation cooling system on line 1RI07B (2)secondary containment operability with the failure of the Unit 2 reactor building exhaust differential pressure controller and Unit 1 reactor building ventilation secured (3)failure of the Unit 1 'A' residual heat removal pump breaker to close during integrated response time testing (4)loose fasteners discovered on the floating end of the cooling water heat exchanger for the Division 1 diesel generator

(5) Unit 2 Division 2 'B' air compressor check valve failed drop test

71111.18 - Plant Modifications

Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (2 Samples)

The inspectors evaluated the following temporary or permanent modifications:

(1) Engineering Change 392059, "Temporary Switched Jumpers to Defeat Shutdown Cooling High Flow and High Reactor Pressure Isolation in Modes 4 and 5"

(2) Engineering Change 402303, "1VY03A Cooling Coil Replacement," and Engineering Change 622556, "2VY03A Cooling Coil Replacement"

71111.19 - Post-Maintenance Testing

Post-Maintenance Test Sample (IP Section 03.01) (13 Samples)

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

(1) Unit 1 'A' residual heat removal system operability and in-service test following system maintenance on February 19, 2020

(2) Unit 1 'A' residual heat removal and reactor core isolation cooling room cooler post-maintenance testing following replacements under Work Orders 4677536 and

===4677537

(3) Unit 1 Division 1 safety bus, 141Y, restoration post-maintenance testing under Work Order 1910789

(4) Unit 1 reactor core isolation cooling pump operability and in-service test on February 28, 2020

(5) Unit 1 reactor core isolation cooling turbine exhaust vacuum breaker replacement and post-maintenance testing on February 22, 2020

(6) Unit 1 scram time testing after control rod drive mechanism and control rod channel replacements on February 25, 2020

(7) Unit 1 standby liquid control system post-maintenance testing after maintenance on February 20, 2020

(8) Unit 1 Division 3 diesel generator after maintenance activities on February 14, 2020

(9) Unit 1 Division 1 integrated response time testing following common diesel generator maintenance activities on February 21, 2020

(10) Unit 1 Division 2 integrated response time testing following common diesel generator maintenance activities on February 22, 2020

(11) Unit 1 safety relief valve testing following replacement under Work Order 4765561

(12) Unit 1 residual heat removal low pressure core injection and residual heat removal service water pump testing after maintenance activities on February 19, 2020

(13) Unit 2 Division 2 diesel generator 'B' air compressor check valve after valve replacement on January 17, 2020

71111.20 - Refueling and Other Outage Activities

Refueling/Other Outage Sample (IP Section 03.01)===

Main article: IP 46775

(1) The inspectors evaluated the Unit 1 refueling outage L1R18 activities from February 10, 2020, to February 28, 2020

71111.22 - Surveillance Testing

The inspectors evaluated the following surveillance tests:

Surveillance Tests (other) (IP Section 03.01)

Main article: IP 71111.22

(1) LIS-RI-316, "Unit 1 Cycled Condensate Storage Tank Low Level Reactor Core Isolation Cooling Suction Functional Test," Work Order 4969973

Inservice Testing (IP Section 03.01) (2 Samples)

(1) LOS-RI-Q5, "Unit 2 Reactor Core Isolation Cooling System Pump Operability and In Service Test," Work Order 4987309

(2) LOS-LP-Q1, "Unit 2 Low Pressure Core Spray Operability and In Service Test," Work Order 4990264

Containment Isolation Valve Testing (IP Section 03.01) (2 Samples)

(1) LTS-100-3, "Main Steam Isolation Valve Local Leak Rate Test," on February 10, 2020

(2) LTS-100-24, "Reactor Core Isolation Cooling Steam Supply Isolation Valve Local Leak Rate Test," Work Order

RADIATION SAFETY

71124.01 - Radiological Hazard Assessment and Exposure Controls

Radiological Hazard Assessment (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards

Instructions to Workers (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated radiological protection-related instructions to plant workers

Contamination and Radioactive Material Control (IP Section 03.03) (2 Samples)

The inspectors evaluated licensee processes for monitoring and controlling contamination and radioactive material

(1) Observed the licensee surveying equipment that exited the Unit 1 Drywell

(2) Observed individuals exiting the Radiological Controlled Area (RCA) and the licensee surveying items that individuals had in their possession when exiting the RCA

Radiological Hazards Control and Work Coverage (IP Section 03.04) (4 Samples)

The inspectors evaluated in-plant radiological conditions during facility walkdowns and observation of radiological work activities

(1) Turbine Sandblasting Activities; Radiation Work Permit (RWP) LA-01-20-00805

(2) Reactor Disassembly; RWP LA-01-20-00901

(3) Dry Well Control Rod (CRD) Exchange; RWP LA-01-20-00513

(4) Unit 1 Reactor Building Scaffolding Building Activities; RWP 02-19-00606 High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (3 Samples)

The inspectors evaluated licensee controls of the following High Radiation Areas and Very High Radiation Areas:

(1) Unit 1 Reactor In-service Inspection Activities

(2) Unit 1 Main Steam Isolation Valve Testing

(3) Unit 1 Reactor Building Scaffolding Building Activities

Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section 03.06) (1 Sample)

(1) The inspectors evaluated radiation worker and radiation protection technician performance as it pertains to radiation protection requirements

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01)===

Main article: IP 71151

(1) Unit 1 January 1, 2019 to December 31, 2019

(2) Unit 2 January 1, 2019 to December 31, 2019

IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (2 Samples)

(1) Unit 1 January 1, 2019 to December 31, 2019

(2) Unit 2 January 1, 2019 to December 31, 2019

IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (2 Samples)

(1) Unit 1 January 1, 2019 to December 31, 2019

(2) Unit 2 January 1, 2019 to December 31, 2019

71152 - Problem Identification and Resolution

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) Action Request 04265960, Large steam leak found downstream of 1G33-F040

71153 - Follow-up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)

Main article: IP 71153

The inspectors evaluated the following licensee event reports (LERs):

(1) LER 2019-001-00, Turbine Stop Valve Limit Switch Failure due to Lubricant Degradation (ADAMS Accession: ML19015A101). 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 did however identify a violation of NRC requirements.

The circumstances surrounding this LER are documented in the results section of this report.

INSPECTION RESULTS

Unresolved Item (Closed)

Main Steam Isolation Valve Fast Stoke Time Test URI 05000373,05000374/2019012-01 71111.21M

Description:

As part of a previous Design Basis Assurance Inspection (DBAI) performed in 2019, the inspectors identified an Unresolved Item concerning the licensees inboard and outboard main steam isolation valves (MSIVs) fast closure time testing. This test is required by Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.3.6 and verifies the MSIVs will stroke within the time frame assumed in the Updated Final Safety Analysis Report (UFSAR) Chapter 15 accident analysis.

Following the DBAI, the licensee provided historical valve testing information which showed the effects of flow through the main steam lines and differences in valve actuator air pressure had been considered as part of demonstrating the main steam isolation valves would close within 3 to 5 seconds.

No findings were identified.

Corrective Action Reference(s): Action Request 04269959, NRC ID - MSIV Stroke Time Design Basis Question, dated August 6, 2019

Primary Containment Isolation Valve was not Properly Qualified for Submergence Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity

Green NCV 05000373/2020001-01 Open/Closed

None (NPP)71152 The inspectors identified a finding of very low safety significance and an associated non-cited violation (NCV) of 10 CFR 50.49, "Environmental qualification of electric equipment important to safety at nuclear power plants," for the failure of the program for qualifying the electric equipment important to safety to prevent submergence. Specifically, during a steam leak, water filled the motor, limit switch compartment, flex conduit, and junction box for a primary containment isolation valve, resulting in the valve becoming inoperable due to submergence.

Description:

On July 21, 2019, Unit 1 lost control room position indication for 1G33-F040, a safety-related primary containment isolation valve that isolates the reactor water cleanup (RWCU) return line from the feedwater system. Upon further investigation, the licensee examined the 1G33-F040 breaker and found the control power fuses blown. After an attempt to replace the fuses was not successful (repeated failure of the fuses), the licensee declared 1G33-F040 inoperable. Subsequently, the licensee secured the RWCU system in accordance with station procedures and began troubleshooting the cause of the blown fuses. As part of the troubleshooting, the licensee sent operators into the outboard MSIV room and discovered a steam leak in the vicinity of 1G33-F040 emanating from a weld in a RWCU piping elbow downstream of the valve. Operators isolated the leak using manual isolation valves located in the outboard MSIV room.

During examination of the valve, the licensee discovered approximately 3.5 gallons of water in the limit switch compartment, flex conduit, and junction box for 1G33-F040. This was documented in the corrective action program as Action Request 4266338. This condition resulted in the submergence of the valve actuator internal electrical components. The licensee determined that the valve control power transformer at the motor control center had shorted to ground and required replacement. Additionally, the valve actuator motor was unable to pass a megger test due to wetting and had to be replaced.

Discussions between the inspectors and the licensee revealed the following. The licensee had concluded that 1G33-F040 did not need to be qualified for submergence due to being located above flood level and was not required to have a motor 'T' drain installed. The licensee stated in Action Request 4288344 that no weep holes were installed in junction boxes associated with safety-related equipment in the outboard MSIV room and none were required because no terminal strips or splices were present in the junction box. Valve 1G33-F040 was part of the list of licensee's equipment covered under 10 CFR 50.49.

The inspectors reviewed licensee requirements for environmental qualification (EQ) of safety-related valves located in the reactor building in the UFSAR and the licensee's EQ binder.

The licensee's UFSAR, Section 3.11, "Environmental Qualification of Electrical Equipment,"

provides the environmental conditions and design bases of safety-related equipment and components at LaSalle County Station. The licensee's UFSAR, Section 3.11.3, "Qualification Results," states, "The results of the EQ for each type of safety-related equipment identified in the LaSalle controlled computer database are included in the extensive file of EQ binders created and maintained for LaSalle County Station."

The inspectors identified an installation requirement in EQ-LS079 that stated the conduit system be oriented such that there is no potential for condensate migration into the limit switch compartment through the conduit. The inspectors reviewed the licensee's implementing procedures for the installation requirements of EQ-LS079 and identified that Volume 3B NSWP-E-03, "Conduit and Junction Box Installation and Inspection," Section 6.12.24, states that weep holes are required for safety-related junction boxes installed in harsh environmentally qualified zones which the outboard MSIV room was designated.

The inspectors reviewed relevant operating experience and presented the operational experience contained in Information Notice 89-63 to the licensee. The information notice stated, in part, "information in this notice is being provided to alert addressees that electrical circuits located above the plant flood level within electrical enclosures may become submerged in water because appropriate drainage has not been provided. Failure of electrical circuits during service conditions, including postulated accidents, can occur due to submergence if water enters these enclosures and there is no provision for drainage. The electrical enclosures addressed by this notice include terminal boxes, junction boxes, pull boxes, conduits, condulets, and other enclosures for end-use equipment (such as limit switches, motor operators, and electrical penetrations), the contents of which may include cables, terminal blocks, electrical splices and connectors. It is expected that recipients will review this information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems."

The inspectors requested video footage obtained by the licensee during the MSIV room entry and noted water from the steam leak condensing in the upper portions of the room, flowing down the MSIV room walls, and over the junction box for 1G33-F040. As such, valve 1G33-F040 was installed in a location in the outboard MSIV room that subjected it to submergence due to condensate migration through its conduit, a condition prohibited by EQ-LS079 under conditions described in Information Notice 89-63.

The inspectors reviewed 10 CFR 50.49, "Environmental qualification of electric equipment important to safety for nuclear power plants," and noted the following:

Title 10 CFR 50.49

(a) stated, in part, each shall establish a program for qualifying the electric equipment defined in paragraph

(b) of this section.

Title 10 CFR 50.49

(b) stated, in part, electric equipment important to safety covered is safety-related electric equipment is that relied upon to remain functional during and following design basis events. Design basis events are defined as conditions of normal operation, including anticipated operational occurrences, design basis accidents, external events, and natural phenomena.

Title 10 CFR 50.49

(e) stated, in part, the electric equipment qualification program must include and be based on submergence (if subject to being submerged).

Based upon the above information, the inspectors determined that the licensee's program for qualifying electric equipment had sufficient information to prevent submergence of valve 1G33-F040 and was within their ability to foresee and correct; however, the information was not used to prevent the valve from failing under submergence.

Additionally, the inspectors reviewed Action Requests AR 04265912, 1G33-F040 Lost Position Indication, AR 04265960, Large Steam Leak Downstream of 1G33-F040, AR 04266338, "Discovery, work performed, lessons learned," and AR 04288344, "NRC question on 1G33-F040," for the following performance attributes:

complete and accurate identification of the problem in a timely manner commensurate with its safety significance and ease of discovery;

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

evaluation and disposition of operability/functionality/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; and

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

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 with the exception that it was the inspectors review and questions surrounding the issue that identified the need for the licensee to look more into the environmental qualifications of the effected valve. The licensee is taking action on improvement in this area.

Corrective Actions: The licensee repaired the valve and completed the appropriate post-maintenance testing including valve diagnostic testing on 1G33-F040 under Work Order 4943440 on July 23, 2019. After all repairs were completed on the RWCU system, the system was restored to operation on July 29, 2019. The licensee captured the inspectors' concerns regarding EQ requirements in the corrective action program for further evaluation.

Corrective Action References: Action Requests 4265960, 4265912, and 4288344

Performance Assessment:

Performance Deficiency: The inspectors determined that the licensee's program for qualifying electric equipment failed to qualify 1G33-F040, RWCU feedwater isolation valve, for submergence as required by 10 CFR 50.49 and was therefore a performance deficiency.

Specifically, when a RWCU through wall leak occurred in the room where the valve was located, the motor controller became submerged, making the valve inoperable and unable to perform its safety function.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance 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 submergence of 1G33-F040, a primary containment isolation valve, impacted the valve ability to close on an isolation signal and prevent radioactive releases to the environment. The inspectors reviewed Inspection Manual Chapter 0612, Appendix E, "Examples of minor issues," and determined there were no applicable examples.

Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. 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 Barrier Integrity questions and determined that the finding was of very low safety significance (Green) since it did not represent an actual open pathway in the physical integrity of the reactor containment, containment isolation system and heat removal components or a reduction of hydrogen igniters in the reactor containment.

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

Enforcement:

Violation: Title 10 CFR Part 50.49(a), "Environmental qualification of electrical equipment important to safety for nuclear power plants," states, in part, the licensee shall establish a program for qualifying the electric equipment defined in paragraph

(b) of this section.

Title 10 CFR Part 50.49(b), states, in part, electric equipment important to safety covered is safety-related electric equipment that is relied upon to remain functional during and following design basis events. Design basis events are defined as conditions of normal operation, including anticipated operational occurrences, design basis accidents, external events, and natural phenomena.

Title 10 CFR Part 50.49(e)(6), states, in part, the electrical equipment qualification program must include and be based on submergence (if subject to being submerged).

Contrary to the above, on July 21, 2019, electrical equipment important to safety in the electrical equipment qualification program did not and was not based on submergence.

Specifically, 1G33-F040, RWCU to feedwater isolation valve, failed due to the motor operator being submerged during a steam leak in the room where the valve was located.

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

Condition Prohibited by Technical Specifications Due to Turbine Stop Valve Limit Switch Failure Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000373/2020001-02 Open/Closed

Not Applicable 71153 The inspectors reviewed a self-revealed severity level IV (SLIV) NCV of 10 CFR 50.36(c)(2)(i), Limiting Conditions for Operation [LCO], for failing to meet/follow the required actions for LCO 3.3.1.1, Reactor Protection System (RPS) Instrumentation, and 3.3.4.1, End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation. Specifically, operators performed surveillance testing per station procedure LOS-RP-Q2 on May 20, 2018, and subsequently on September 9, 2018, during which turbine stop valve limit switch #3 was found inoperable. The licensee later determined after a failure analysis of the limit switch that during the May 20, 2018, surveillance that turbine stop valve #3 limit switch, 1C71-N006C, was actually failed rendering the associated RPS and EOC-RPT channels inoperable on May 20, 2018, even though corrective actions were taken and the limit switch was declared operable at that time.

Description:

On May 20, 2018, while performing surveillance testing on Unit 1 turbine stop valve (TSV) #3 per station procedure LOS-RP-Q2, the 1C71A-K10C relay (reactor protection input to the 'A' channel) took approximately 2 to 6 minutes to de-energize, but the 1C71A-K10F relay (reactor protection input to the 'B' channel) de-energized as expected. The TSV otherwise functioned normally, by closing and fast closing as expected. Further investigation determined that the closure RPS channel A2 scram limit switch (1C71-N006C) that actuates relay 1C71A-K10C was stuck closed. The relay was declared inoperable and the appropriate Technical Specifications (TS) were entered. This was documented in the corrective action program in Action Request 4139358.

Troubleshooting was performed under Work Order 4684002 during which it was observed that the limit switch actuating arm failed to return to the relaxed position following movement of the TSV actuating rod assembly. The licensee applied additional manufacture's lubricant to the limit switch and then exercised the limit switch. The surveillance test was performed again with the limit switch and relay performing satisfactorily. The relay was declared operable and the applicable TS were exited.

On September 9, 2018, the next performance of surveillance testing on the Unit 1 TSV #3 per station procedure LOS-RP-Q2, the 1C71A-K10C relay failed to de-energize when the valve closed during testing. The relay was declared inoperable and applicable TS were entered.

Work Order 1466840 was completed to replace the TSV #3 1C71-N006C limit switch.

Testing was performed satisfactorily, and the relay was declared operable and the applicable TS were exited. This was documented in the corrective action program as Action Request 4171148.

The limit switch was sent off for failure analysis that was completed on November 16, 2018.

The analysis determined that the cause of the failure was degradation of the switch lubricant due to exposure to a high temperature environment. The report stated that this fact established firm evidence that the switch performance in May 2018 was related to the failure of the switch in September 2018. The licensee performed a review of the occurrence and determined that the switch had been inoperable since the occurrence of the switch malfunction on May 20, 2018, and required reporting in accordance with 10 Code of Federal Regulations 50.73(a)(2)(i)(B) as a condition prohibited by the plant's TS.

Licensee Event Report (LER) 2019-001-00, "Turbine Stop Valve Limit Switch Failure Due to Lubricant Degradation," was reported to the NRC on January 15, 2019, and stated that the switch had been inoperable from the first occurrence of the switch malfunction on May 20, 2018, until corrective action was taken in response to the surveillance test failure on September 9, 2018, based on firm evidence of the cause of the malfunction. The inoperability period was greater than allowed by TS LCO 3.3.3.1 and TS LCO 3.3.4.1 as follows:

Limiting Condition for Operation 3.3.1.1 required that the RPS instrumentation for each function is OPERABLE, and Condition A applies to one or more required channels being inoperable and required that the channel or its associated trip system is placed in trip within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . In this case, the station had exceeded the completion time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , when the malfunctioning channel was found inoperable during the September 2018 surveillance test.

In addition, LCO 3.3.1.1 Condition D was not entered for the completion time of Condition A not met.

Limiting Condition for Operation 3.3.4.1 required that the end of cycle recirculation pump trip (EOC-RPT) instrumentation is OPERABLE, which included two channels of the TSV closure and turbine control valve (TCV) fast closure or minimum critical power ratio (MCPR) limits are applied. This condition was not met.

The resident inspectors requested and reviewed applicable documentation and licensee records and determined the following.

In 2004, two limit switch failures (1C and 2G) occurred and the licensee performed an equipment apparent cause evaluation that included a failure analysis that determined the cause of the failure was degradation of the lubricant in the limit switch under high temperatures. The licensee could not find limit switches for higher temperature application but changed the preventative maintenance strategy to replacement every 4 years.

In 2009, the replacement strategy preventative maintenance was inadvertently retired when a digital electrohydraulic control modification was done on the main generators for both units.

In 2011, a limit switch (1C) had shown degraded performance during a surveillance test and was replaced. The failure analysis determined degradation of the lubricant at high temperature to be the cause. The preventative maintenance to replace the limit switches was re-implemented at a 6-year frequency. No high temperature application switches were sought after at this time.

In 2014, a limit switch (2G) showed degraded performance during a surveillance test and was lubricated, exercised, and retested satisfactorily. An equipment apparent cause was performed that determined degradation of the lubricant at high temperature was the cause, and that based on environmental qualification data, the switch should be replaced on a 4-year frequency.

In 2017, a limit switch (1C) stuck closed during a surveillance test and was lubricated, exercised, and retested satisfactorily. A work group evaluation was done with no changes to the replacement preventative maintenance.

In January 2018, a limit switch (2E) stuck closed during a surveillance test and was replaced.

The failure analysis determined degradation of the lubricant at high temperature to be the cause. Based on the failure, the replacement frequency was changed to every two years.

In February 2018, all Unit 1 limit switches susceptible to high temperature lubricate degradation were replaced during the refueling outage.

In May 2018, a limit switch (1C) showed degraded performance during a surveillance test and was lubricated, exercised, and retested satisfactorily. The failure analysis determined degradation of the lubricant at high temperature to be the cause. A high temperature application switch had still not been identified. Additional actions were put in place to ensure all Unit 2 switches susceptible to high temperatures would be replaced in the 2019 refueling outage.

In early September 2018, a high temperature application limit switch was identified and approved for use. The limit switches would be installed during the upcoming refueling outages in 2019 for Unit 2 and 2020 for Unit 1.

In mid-September 2018, a limit switch (1C) stuck closed during a surveillance test and was replaced. The failure analysis determined degradation of the lubricant at high temperature to be the cause. The subject LER was written to report the conditions prohibited by the TS.

High temperature application switches have been installed in both units as of February 2020.

Based on the information reviewed by the inspectors, they determined that for the events that occurred in the subject LER in May and September of 2018, that the licensee was taking appropriate action to correct the issue by finding high temperature application switches and scheduling replacement of the switches prior to the failure in mid-September and that no performance deficiency existed; however, because of the limit switch failure, a violation of TS had occurred.

Corrective Actions: The licensee has replaced all effected limit switches on both units with limit switches using grease suitable for the higher temperature.

Corrective Action References: Action Requests 4048633, 4139358, 4171148, and 4171578

Performance Assessment:

The NRC determined this violation was not reasonably foreseeable and preventable by the licensee and therefore is not a performance deficiency. Specifically, for the events that occurred in the subject LER in May 2018 and September 2018, the licensee took appropriate action to correct the issue by finding high temperature application switches and scheduling replacement of the switches prior to the failure in mid-September and that no performance deficiency existed.

Enforcement:

Severity: Traditional Enforcement is being used to disposition this violation with no associated Reactor Oversight Process performance deficiency, per NRC Memorandum Interim Guidance for Dispositioning Severity Level IV Violations with No Associated Performance Deficiency, dated June 15, 2018 (ADAMS Accession No. ML18158A220). The inspectors reviewed this issue in accordance with Inspection Manual Chapter 0612 and the Enforcement Manual. Reactor Violations without a performance deficiency are dispositioned using the traditional enforcement process. The inspectors reviewed Section 6.1.d.1 of the Enforcement Policy and determined this violation was Severity Level IV because it was a failure to comply with a TS action requirement for an LCO in Section 3.0.

Violation: Title 10 CFR 50.36(c)(2)(i), Limiting Conditions for Operation," states, in part, that when a LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TS until the condition can be met.

LaSalle TS LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation," requires that the RPS instrumentation for each function in Table 3.3.1.1-1 shall be OPERABLE.

LaSalle TS LCO 3.3.4.1, "End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation,"

requires that two channels per trip system for each EOC-RPT instrumentation Function listed below shall be OPERABLE:

1. Turbine Stop Valve - Closure: and

2. Turbine Control Valve Fast Closure, Trip Oil Pressure-Low

OR

LaSalle TS LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," limits for inoperable EOC-RPT as specified in the COLR are made applicable.

Contrary to, on May 20, 2018, when an LCO of a nuclear reactor was not met, the licensee failed to shut down the reactor or follow any remedial action permitted by the TS until the condition could be met. Specifically, when the limit switch failed the licensee did not perform remedial actions for LaSalle TS LCO 3.3.1.1 and 3.3.4.1 until the LCO could be met.

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

EXIT MEETINGS AND DEBRIEFS

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

On April 8, 2020, the inspectors presented the integrated inspection results to Mr. P. Hansett, Plant Manager, and other members of the licensee staff.

On March 2, 2020, the inspectors presented the Radiation Protection Inspection, inspection results to Mr. J. Washko, Site Vice President, and other members of the licensee staff.

On April 6, 2020, the inspectors presented the inservice inspection (ISI) baseline inspection results to Mr. P. Hansett, Plant Manager, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

71111.04

Drawings

1RI07B-2-2

Inch and Under As-Built

G

M-147, Sheet 1

P&ID Reactor Core Isolation Coolant System (RCIC)

BL

M-147, Sheet 2

P&ID Reactor Core Isolation Coolant System (RCIC)

AP

Miscellaneous

STI Change

Request Number

LA-18-09

Frequency Change from 24 to 48 Months

Procedures

LOP-RI-02E

Unit 2 Reactor Core Isolation Cooling System Electrical

Checklist

LOP-RI-02M

Unit 2 Reactor Core Isolation Cooling System Mechanical

Checklist

71111.05

Procedures

PFP FZ 4A

Pre-Fire Plan, Auxiliary Building 815' Elevation, Upper

Ventilation Equipment Floor, Fire Zone 4A

PFP FZ 4B

Pre-Fire Plan, Auxiliary Building 768' Elevation, Lower

Ventilation Equipment Floor, Fire Zone 4B

PFP FZ 4E2

Pre-Fire Plan, Auxiliary Building 731' Elevation, Unit 2

Auxiliary Electrical Equipment Room, Fire Zone 4E2

PFP FZ 4E4

Pre-Fire Plan, Auxiliary Building 731' Elevation, Unit 2

Division 2 Essential Switchgear Room, Fire Zone 4E4

PFP FZ 4F2

Pre-Fire Plan, Auxiliary Building 710' Elevation, Unit 2

Division 1 Essential Switchgear Room, Fire Zone 4F2

Work Orders

4575190

Perform Unit 2 Fire Zone Ionization Smoke Detector

Channel Functional Test, LES-FP-05, Attachment 2F, Zone

2-9

06/06/2018

4678651

Fire Damper Visual Inspection, LMS-FP-22, Attachment I

11/08/2019

4719503

Fire Damper Visual Inspection, LMS-FP-22, Attachment K

11/19/2019

71111.07A

Procedures

ER-AA-340-1002,

1, 1A

RHR Heat

Exchanger

1A RHRSW Heat Exchanger (1E12-B001A) GL 89-13

Inspection Report

2/21/2020

71111.08G Corrective Action

Documents

AR 04108612

VT-3 Rejectable Indications on 1E51-F063 Valve

2/27/2018

AR 04110334

VT-3 Rejectable Indication on HG06-1088G

03/02/2018

AR 04110843

Foreign Material Exclusion: In-Vessel Visual Inspection

03/03/2018

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

Discovered Foreign Material Within Reactor Pressure Vessel

AR 04111184

Foreign Material Exclusion: Discovered Within Reactor

Pressure Vessel

03/04/2018

AR 04112887

Foreign Material Exclusion: Discovered Shroud Head Bolt

Missing Anti-Rotation Pin

03/08/2018

AR 04125563

1E12-F068A Valve Yoke/Bolting and Packing Area

Corrosion

04/11/2018

AR 04265960

Large Steam Leak Found Downstream of 1G33-F040

07/22/2019

AR 04286478

Isolated Pit Identified on Core Standby Cooling System

Discharge Pipe 1RH83BA-24

10/29/2019

AR 04316736

Visual Exam Results for Inservice Inspection Component,

High Pressure Core Spray Diesel Generator Cooler 1

2/10/2020

AR 04316738

Visual Exam Results for Inservice Inspection Component,

High Pressure Core Spray Pump

2/10/2020

Corrective Action

Documents

Resulting from

Inspection

AR 04318890

NRC Question Regarding Code Reconciliation

2/18/2020

Drawings

E-232-861

Vessel Support Skirt Assembly and Details

IT-7000-M-PS-06

Installation Tolerances - Pipe Supports

B

IT-7000-M-PS-07

Installation Tolerances - Pipe Supports

C

IT-7000-M-PS-09

Installation Tolerances - Pipe Supports

C

M09-HG06-

1088R

Unit 1 Pipe Support HG06-1088R

A

M09-RR00-

1013S

Unit 1 Pipe Support RR00-1013S

D

M09-RR00-

1015S

Unit 1 Pipe Support RR00-1015S

C

Engineering

Changes

EC 390160

LaSalle Units 1 and 2, American Society of Mechanical

Engineers Code Reconciliation and Applicability of Later

Code Editions and Addenda to Section XI and Non-Section

XI Activities for Revised Allowable Stress Values

71111.08G Engineering

Evaluations

EC 629614

Evaluation of Non-Destructive Examination Report 19-211

for Line 1RH83BA-24

Miscellaneous

10CFR50.59

Screening L12-

151

American Society of Mechanical Engineers Code

Reconciliation and Applicability of Later Code Editions and

Addenda to Section XI and Non-Section XI Activities for

Revised Allowable Stress Values

RRP 2635

Repair/Replacement Plan: Remove Degraded Piping and

Degree Elbow Downstream of 1G33-F040 Valve and

Replace with New

07/23/2019

NDE Reports19-163

Radiographic Examination Interpretation Report: Pipe to 90

Degree Elbow to Pipe, Weld 2 and Weld 3

07/24/2019

19-164

Magnetic Particle Examination Report: Piping Weld 2 and

Weld 3

07/24/2019

19-174

Radiographic Examination Interpretation Report: Piping

Weld W-1A

07/28/2019

19-175

Radiographic Examination Interpretation Report: Piping

Weld W-4

07/28/2019

19-176

Radiographic Examination Interpretation Report: Piping

Weld W-5

07/28/2019

19-177

Magnetic Particle Examination Report: Piping Weld R4

Excavation

07/28/2019

19-178

Radiographic Examination Interpretation Report: Piping

Weld W-R4

07/29/2019

19-183

Magnetic Particle Examination Report: Piping Weld 1A,

Weld 4/R4, and Weld 5

07/29/2019

2018-PT-059

Liquid Penetrant Examination Report: Valve Internal / Disc

Trunnion - Component ID 1E51-F063

2/27/2018

2018-VT-054

Visual Exam of Pumps and Valves (VT-3) Report: Disk

SN-7U, Component ID 1E51-F063

2/26/2018

2018-VT-056

Visual Exam of Pumps and Valves (VT-3) Report: Valve

Internals New and Used, Component ID 1E51-F063

2/26/2018

E19-183

Visual Examination (VT-2) Report: Line 1FW11A-4

Replacement Elbow and Pipe Pup Pieces Downstream of

1G33-F040

07/29/2019

L1R17-VT-073

Visual Examination of Support (VT-3) Report: Rigid Support,

Component ID HG06-1088R

03/04/2018

L1R18-MT-001

Magnetic Particle Examination Report: Reactor Vessel Skirt,

Component ID IVS-1-2-3

2/13/2020

L1R18-UT-001