IR 05000373/2024001
| ML25092A242 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 04/03/2025 |
| From: | Robert Ruiz NRC/RGN-III/DORS/RPB1 |
| To: | Rhoades D Constellation Energy Generation |
| Shared Package | |
| ML25091A196 | List: |
| References | |
| IR 2024001 | |
| Download: ML25092A242 (1) | |
Text
SUBJECT:
REISSUE LASALLE COUNTY STATION - INTEGRATED INSPECTION REPORT 05000373/2024001 AND 05000374/2024001
Dear David P. Rhoades:
The U.S. Nuclear Regulatory Commission (NRC) is re-issuing this inspection report in its entirety due to revising a non-cited violation (NCV) issued in Inspection Report 05000373;05000374/2024001 dated May 13, 2024 (Agencywide Document and Access Management System (ADAMS) Accession No. ML24131A151). The NRCs revision is a result of the agencys review of your letter dated June 12, 2024 (ML24164A061), in which you contested an NCV associated with a failure to test a motor operated valve following maintenance as required by your inservice testing program (see Section 71111.24 of the Enclosure to this letter). The details associated with our review and the NCV revision were provided to you in an NRC letter dated April 2, 2025 (ML25076A656).
If you contest the revised violation or the significance or severity of the revised violation documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement; and the NRC Resident Inspector at LaSalle County Station.
Since no changes were made to any of the cross-cutting aspects assigned in the original report and more than 30 days have elapsed since the reports original issuance date of May 13, 2024, the opportunity to disagree with the cross-cutting aspects assigned in this report has expired.
April 3, 2025 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, Robert Ruiz, Chief Reactor Projects Branch 1 Division of Operating Reactor Safety Docket Nos. 05000373 and 05000374 License Nos. NPF-11 and NPF-18 Enclosure:
As stated cc w/ encl: Distribution via LISTSERV Stoedter, Karla signing on behalf of Ruiz, Robert on 04/03/25
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 Follow Seismic Storage Requirements for a Temporary Battery Charger Stacked on an Unrestrained Cart Configuration Adjacent to the Safety-Related Unit 1 125Vdc Division 2 Battery Rack and Exposed Terminals Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000373/2024001-01 Open/Closed
[H.8] -
Procedure Adherence 71111.05 The inspectors identified a Green finding and an associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when the licensee failed to follow the requirements of station procedure LAP-100-56, Equipment Parts Storage in Plant Areas Containing Safety-Related Equipment, Revision 10. Specifically, the licensee failed to ensure that a portable battery charger stacked on the top of a mobile cart with a combined height to width ratio of greater than 2.0 was either stored in an approved seismic storage area, seismically restrained, approved by engineering, or stored at least a distance of 2 feet plus the height of the stacked configuration away from the safety-related batteries with exposed terminals in accordance with the procedure requirements.
Failure to Test Motor-Operated Valve in Accordance with the Inservice Test Program Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000374/2024001-02 Open/Closed
[H.11] -
Challenge the Unknown 71111.24 The inspectors identified a finding of very low safety significance (Green) and a non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50.55a(f)(4)(ii)for the licensees failure to meet the in-service testing requirements set forth in the American Society of Mechanical Engineers (ASME) Operations and Maintenance Code and Addenda Code Case OMN-1 after performing maintenance that could affect motor-operated valve (MOV) performance. Specifically, the licensee failed to perform testing on primary containment isolation MOV, 2B21-F016 a valve within the scope of the ASME OM Code and Addenda, before returning the valve to service after performing a maintenance activity that could affect the valves performance. The maintenance activity involved disconnecting and reconnecting the MOVs circuitry as part of installing and removing an MOV backseat relay tool used to electrically backseat the valve.
Failure to Use Calibrated Measuring and Test Equipment to Electrically Backseat a Safety Related Motor-Operated Valve Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000374/2024001-03 Open/Closed
[H.11] -
Challenge the Unknown 71111.24 The inspectors identified a finding of very low safety significance (Green) and a NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XII,
Control of Measuring and Test Equipment, for the licensees failure to assure a tool used in activities affecting quality was properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits. Specifically, the licensee failed to apply quality assurance requirements to the MOV BSRT when using the tool to electrically backseat a Unit 2 safety-related valve.
Failure to Promptly Correct Degraded Pressure Switches in the Unit 1 and Unit 2 Main Steam Line High Flow Isolation Logic System Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000373,05000374/2024001-04 Open/Closed None (NPP)71152A The inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly correct a condition adverse to quality associated with the Unit 1 and Unit 2 main steam line (MSL) isolation logic system. Specifically, the licensee identified in a 2008 equipment apparent cause evaluation that pressure switches installed in the Unit 1 and Unit 2 MSL high-flow isolation logic trip systems were susceptible to multiple failure modes and had been exposed to peak inductive currents during frequent calibration activities that may have damaged switch contacts. A subsequent corrective action replaced half of the impacted switches while the other half continued to be replaced on an as-needed basis.
Additional Tracking Items
Type Issue Number Title Report Section Status URI 05000373,05000374/20 23002-01 Clarification of the Dew Point Specification for the MSA Firehawk M7 SCBA System 71124.03 Closed
PLANT STATUS
Unit 1 began the inspection period at rated thermal power. On February 19, 2024, the unit coasted down to approximately 86 percent power and commenced refueling outage L1R20.
On March 8, 2024, the unit started up and reached rated thermal power on March 13, 2024.
On March 17, 2024, the unit was down powered to approximately 60 percent to perform power suppression testing for a suspected fuel cladding failure. The suspected fuel failure was suppressed and later that day, the unit returned to rated thermal power and remained at or near rated thermal power for the duration of the inspection period.
Unit 2 began the inspection period at rated thermal power. On January 12, 2024, the unit was down powered to approximately 80 percent for a control rod sequence exchange. The unit was returned to rated thermal power on the same day and remained at 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 performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.01 - Adverse Weather Protection
Seasonal Extreme Weather Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated the licensees readiness for extreme cold weather condition preparation for the following systems:
1. unit common A train firewater pump
2. cooling water intake systems
3. FLEX buildings
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)1A residual heat removal system on January 4, 2024, after quarterly pump run
- (2) Unit 1 and Unit 2 station air following the trip of a station air compressor on March 26, 2024
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (9 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 312, reactor building, elevation 673'-4", Unit 2 high-pressure core spray cubicle on January 17, 2024
- (2) Fire Zone 3H2, reactor building, elevation 694'-6", Unit 2 high-pressure core spray cubicle on January 19, 2024
- (3) Fire Zone 2K, reactor building, elevation 687'-0" to 768'-0", Unit 1 steam tunnel on February 22, 2024
- (4) Unit 2 auxiliary electrical equipment room on March 13, 2024
- (5) Fire Zone 3B1, reactor building, elevation 820'-6", Unit 2 general area and standby gas treatment area on March 13, 2024
- (6) Fire Zone 2B1, reactor building, elevation 820'-6", Unit 1 general area and standby gas treatment area on March 13, 2024
- (7) Fire Zone 5D3, elevation 687'-0", Unit 1 high-pressure core spray switchgear area on March 14, 2024
- (8) Fire Zone 5D2, elevation 687'-0", Unit 2 high-pressure core spray switchgear area on March 14, 2024
- (9) Fire Zone 3I4, reactor building, Unit 2 low-pressure coolant system (LPCS)/reactor core isolation cooling (RCIC) pump cubicle room on March 29, 2024
71111.06 - Flood Protection Measures
Flooding Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated the temporary floor drain plugging configuration in the turbine building to support the movement of 250Vdc batteries (RE: Engineering Change 640690).
71111.08G - Inservice Inspection Activities (BWR)
BWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding
Activities (IP Section 03.01)
The inspectors evaluated boiling-water reactor nondestructive testing by reviewing the following examinations from February 26-March 1, 2024:
(1)
1. visual examination of residual heat removal (RHR) constant support
RH40-1004C
2. visual examination of MS restraint MS01-1352X
3. ultrasonic examination of reactor pressure vessel (RPV) recirculation outlet
nozzle inner radius 1-NIR-10
4. ultrasonic examination of RPV core spray nozzle inner radius 1-NIR-16
5. ultrasonic examination of RPV bottom head meridional welds GEL-1006 DB,
6. weld repair on Unit 1 RHR HX weld 1RH-HX1B-9A
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 reactor startup from outage L1R20 on March 8, 2024.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated license operator requalification simulator out-of-box evaluation on March 19, 2024.
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 2 elevated action Green risk due to main steam line high-flow channel spurious trips on January 17, 2024
- (2) Unit 2 L1R20 shutdown safety plan
- (3) Unit 2 planned elevated action Green risk for behind-the-meter related activities on February 1, 2024
- (4) Unit 1 planned reduced inventory to support head removal on February 20, 2024
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (8 Samples)
The inspectors evaluated the licensees justifications and actions associated with the following operability determinations and functionality assessments:
- (1) Unit 2 channel A1 main steam isolation valve high-flow isolation operability between spurious signals on January 9, 2024, and January 12, 2024 (2)operability assessment of the Unit 2B emergency diesel generator after small amounts of metal debris were observed under the top deck cover on January 23, 2024
- (3) Unit 2 RCIC issues with manual control
- (4) Unit 2 main stem line valve drain inboard isolation valve 2B21-F016 back seating activity to reduce suspected packing leakage
- (5) Unit 1 1B reactor recirculation seal pressure temperature and pressure fluctuations
- (6) Unit 1 1A reactor recirculation seal pressure temperature and pressure fluctuations
- (7) Unit 1 L1R20 water rod inspection identification of three fuel spacers that shifted up near the upper core plate (8)2CM023B operability as a primary containment isolation valve and associated containment monitoring channel operability given continued indication and manufacturing issues
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)
- (1) Unit 2 main steam inboard drain primary containment isolation valve backseat to reduce suspected packing leakage
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated Unit 1 refueling outage L1R20 related activities from February 19, 2024, to March 8, 2024.
71111.24 - Testing and Maintenance of Equipment Important to Risk
The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (4 Samples)
- (1) Unit 1 division 2 safety-related battery cell 56 equalizing charge on January 31, 2024
- (3) Unit 1 reactor vessel leakage test on March 5, 2024
- (4) Unit 1 1A emergency diesel generator fast start on March 29, 2024
Surveillance Testing (IP Section 03.01) (5 Samples)
- (1) Unit 2 main steam line high-flow main steam isolation valve (MSIV) isolation calibration per LIS-MS-202 on January 9, 2024
- (2) Unit 1 MSIV stroke time testing on February 19, 2024
- (3) Unit 1 reactor recirculation flow control valve lockup testing on February 19, 2024
- (4) Unit 1 integrated division 1 response time surveillance per LOS-DG-109 on February 21, 2024
- (5) Unit 2 2A emergency diesel generator idle start testing on March 15, 2024
Inservice Testing (IST) (IP Section 03.01) (2 Samples)
- (1) Unit 2 2A standby liquid control pump on March 15, 2024
- (2) Unit 2 2B emergency diesel generator cooling water pump comprehensive inservice test on March 18,
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, 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 how the licensee instructs workers on plant-related radiological hazards and the radiation protection requirements intended to protect workers from those hazards.
Contamination and Radioactive Material Control (IP Section 03.03) (2 Samples)
The inspectors observed/evaluated the following licensee processes for monitoring and controlling contamination and radioactive material:
(1)licensee surveys and decontamination of potentially contaminated material leaving the main radiologically controlled area (2)licensee surveys and decontamination of potentially contaminated material leaving the radiologically controlled area at the north service building
Radiological Hazards Control and Work Coverage (IP Section 03.04) (4 Samples)
The inspectors evaluated the licensees control of radiological hazards for the following radiological work:
- (1) Unit 1 undervessel installation
- (2) Unit 1 high-pressure turbine diaphragm inspection (3)refueling floor cavity platform activities
- (4) Unit 1 F004 valve room activities 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 building personnel access
- (2) Unit 1 turbine deck center court
- (3) Unit 1 drywell entry 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.
71124.03 - In-Plant Airborne Radioactivity Control and Mitigation
Temporary Ventilation Systems (IP Section 03.02) (1 Sample)
The inspectors evaluated the configuration of the following temporary ventilation systems:
- (1) LAS-416 HEPA unit
Use of Respiratory Protection Devices (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated the licensees use of respiratory protection devices.
71124.08 - Radioactive Solid Waste Processing & Radioactive Material Handling,
Storage, & Transportation
Shipment Preparation (IP Section 03.04)
- (1) The inspectors observed the preparation of radioactive shipment LM24-003 of control rod drive boxes on February 29,
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)===
- (1) Unit 1 (January 1, 2023, through December 31, 2023)
- (2) Unit 2 (January 1, 2023, through December 31, 2023)
IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (2 Samples)
- (1) Unit 1 (January 1, 2023, through December 31, 2023)
- (2) Unit 2 (January 1, 2023, through December 31, 2023)
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (2 Samples)
- (1) Unit 1 (January 1, 2023, through December 31, 2023)
- (2) Unit 2 (January 1, 2023, through December 31, 2023)
71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) The inspectors evaluated spurious Unit 2 channel A1 main steam line isolation valve isolation signals that occurred on several dates as documented in Action Requests (ARs) 4730751 and 4727857. A finding of very low safety significance (Green) and non-cited violation is associated with the inspectors review and are documented in the Inspection Results section of this report.
71153 - Follow-Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 05000374/2023-003-00, LaSalle County Station, Unit 2, Automatic Actuation of Reactor Protection System (RPS) During Restoration from Hydrostatic Test Conditions (ADAMS Accession No. ML23122A016). The inspectors determined that the cause of the condition described in the LER does not represent a finding because the unit was already shut down at the time of actuation and the RPS signal was not applicable. This LER is closed.
INSPECTION RESULTS
Failure to Follow Seismic Storage Requirements for a Temporary Battery Charge Stacked on an Unrestrained Cart Configuration Adjacent to the Safety-Related Unit 1 125Vdc Division 2 Battery Rack and Exposed Terminals Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000373/2024001-01 Open/Closed
[H.8] -
Procedure Adherence 71111.05 The inspectors identified a Green finding and an associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when the licensee failed to follow the requirements of station procedure LAP-100-56, Equipment Parts Storage in Plant Areas Containing Safety-Related Equipment, Revision 10. Specifically, the licensee failed to ensure that a portable battery charger stacked on the top of a mobile cart with a combined height to width ratio of greater than 2.0 was either stored in an approved seismic storage area, seismically restrained, approved by engineering, or stored at least a distance of 2 feet plus the height of the stacked configuration away from the safety-related batteries with exposed terminals in accordance with the procedure requirements.
Description:
During the walkdown portion of this inspection, the inspectors identified that a portable battery charger stacked on a mobile cart was within in very close proximity to the Unit 1 division 2 125Vdc safety-related batteries. The licensee was using the temporary battery charger for an extended single charge of cell #56 per licensee procedure LEP-DC-01. The inspectors identified a concern that if a seismic event occurred, or if bumped, that the stacked metal frame battery charger unit could fall off the plastic cart and onto the exposed battery terminals resulting in an arc event and loss of 125Vdc battery charger or batteries themselves. The inspectors immediately notified the Unit 1 senior reactor operator of their concerns. The licensee promptly secured the single cell charge, removed the cart and battery charger from the 125Vdc division 2 battery room. The licensee discussed the station storage requirements and expectations with the appropriate site personnel and entered this issue into their corrective action program (AR 4739747). The licensee concluded that this configuration did not meet station requirements and expectations, however, the battery and associated permanent battery charger remained operable in this configuration based upon their judgement.
The inspectors reviewed the licensee corrective action document and procedure that maintains structure, system, and component quality with respect to equipment storage, LAP-100-56, revision 10, Equipment/Parts Storage in Plant Areas Containing Safety-Related Equipment. A specific purpose listed in the procedure was to define the LaSalle station requirements for equipment/parts storage in plant areas containing safety-related equipment. The inspectors approximated the height (H) to weight (W) ratio to be 6 based upon the smallest width of the battery charger combined with the total height of the stacked configuration to be approximately 48 inches. The inspectors approximated the distance from the stacked metal frame battery charge to be approximately 1 foot higher and 10 inches from the open battery terminals. The inspectors identified the specific procedural requirements not met included:
Specifically (Ref: LAP-100-56, revision 10):
B.4.2.1 Stored items with a height (H) to width (W) ration (H/W) greater than 2.0 may require a seismic restraint.
B.4.2.2 If H/W is greater than 2.0, proceed to B.4.4...
B.4.3 Stored items with H/W < 2.0....
B.4.4 Stored items with H/W > 2.0.
B.4.4.1 Items stored inside or outside one of the approved storage areas.
B.4.4.1.1 If the edge of the stored item is located at least 24-inches plus the height (H) of the stored item away from the nearest piece of safety-related equipment/component...
otherwise proceed to B.4.6 (seismic restraint required).
B.4.6 Seismic Restraint B.4.6.1 All stored items not meeting the requirements of B.4.3 and B.4.4 shall require seismic restraints unless otherwise approved by Engineering...
Procedure LAP-100-56, revision 10, attachment, provides an activity flow chart which restates the procedural requirements in a different format.
Corrective Actions: The licensee entered this condition into the stations corrective action program AR 4739747. Corrective actions include promptly securing the single cell charger and relocating the cart and stacked battery charger out of the Unit 1 125Vdc division 2 battery room, individual coaching, and an assignment to determine if the procedure could be enhanced.
Corrective Action References: AR 4739747
Performance Assessment:
Performance Deficiency: The licensees failure to follow station procedure LAP-100-56, Equipment Parts Storage in Plant Areas Containing Safety-Related Equipment, Revision 10, was a performance deficiency. Specifically, the procedure accurately captured the licensee requirements in both the procedure body and associated attachment A and therefore was reasonable for the licensee to foresee and prevent.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration 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, one of the purposes of licensee procedure LAP-100-56 is to ensure safety-related components such as the Unit 1 division 2 125Vdc battery remain available, reliable, and capable, during design-basis events. The cart and stacked battery charger were in an unrestrained configuration such that if a seismic event occurred, the battery charger could have tipped forward onto and shorted out 125Vdc battery terminal(s)resulting in the loss of Unit 1 division 2 125Vdc safety-related batteries and/or charger.
Absent a seismic event, the charger could have also been knocked over onto the safety-related battery terminals during normal day-to-day plant operations. In addition, the inspectors reviewed the more-than-minor examples in NRC Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues. The inspectors informed their use of the more-than-minor questions by comparing this finding to the more-than-minor example 3a.
This example illustrates a calculational error with the potential to adversely affect the mitigating system cornerstone objective. Similar to example 3a, the inspectors determined that the finding was more than minor, regardless of the licensees operability assessment, based upon the inspectors reasonable doubt that this stacked configuration could have adversely effected Unit 1 division 2 125Vdc battery availability, reliability, and capability had the battery charger fallen off the mobile cart during a seismic event.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The finding is a deficiency affecting the qualification of a mitigating SSC that maintained its operability.
Cross-Cutting Aspect: H.8 - Procedure Adherence: Individuals follow processes, procedures, and work instructions. Specifically, the inspectors determined that a primary cause for the performance deficiency was because the licensee did not follow procedure LAP-100-56, Equipment Parts Storage in Plant Areas Containing Safety-Related Equipment, Revision 10, as part of this work activity.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings requires, in part, that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures. The licensee established LAP-100-56, Equipment Parts Storage in Plant Areas Containing Safety-Related Equipment, Revision 10, as the implementing procedure for storing equipment near safety-related equipment, as an activity affecting quality.
Procedure LAP-100-56, Revision 10, states:
B.4.2.1 Stored items with a height (H) to width (W) ration (H/W) greater than 2.0 may require a seismic restraint.
B.4.2.2 If H/W is greater than 2.0, proceed to B.4.4...
B.4.3 Stored items with H/W < 2.0....
B.4.4 Stored items with H/W > 2.0.
B.4.4.1 Items stored inside or outside one of the approved storage areas.
B.4.4.1.1 If the edge of the stored item is located at least 24-inches plus the height (H) of the stored item away from the nearest piece of safety-related equipment/component...
otherwise proceed to B.4.6 (seismic restraint required).
B.4.6 Seismic Restraint B.4.6.1 All stored items not meeting the requirements of B.4.3 and B.4.4 shall require seismic restraints unless otherwise approved by Engineering...
Contrary to the above, from January 29, 2024, to January 31, 2024, the licensee failed to follow step B.4.6.1 of procedure LAP-100-56, Revision 10. Specifically, the licensee failed to use required seismic restraints or have the storage configuration approved by Engineering when a temporary battery charger was stored in a stacked configuration on top of a mobile cart with an overall height to width ratio of approximately 6 within approximately 10 inches of the safety-related Unit 1 division 2 125Vdc batteries.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Test Motor-Operated Valve in Accordance with the Inservice Test Program Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000374/2024001-02 Open/Closed
[H.11] -
Challenge the Unknown 71111.24 The inspectors identified a finding of very low safety significance (Green) and a non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50.55a(f)(4)(ii)for the licensees failure to meet the in-service testing requirements set forth in the American Society of Mechanical Engineers (ASME) Operations and Maintenance Code and Addenda Code Case OMN-1 after performing maintenance that could affect motor-operated valve (MOV) performance. Specifically, the licensee failed to perform testing on primary containment isolation MOV, 2B21-F016, a valve within the scope of the ASME OM Code and Addenda, before returning the valve to service after performing a maintenance activity that could affect the valves performance. The maintenance activity involved disconnecting and reconnecting the MOVs circuitry as part of installing and removing an MOV backseat relay tool used to electrically backseat the valve.
Description:
In December 2023, the licensee noticed an increasing trend in Unit 2 reactor coolant system (RCS) unidentified leakage inside the drywell. The licensee performed troubleshooting to identify the possible leak sources and identified a potential packing leak of the main steam isolation valve (MSIV) drain header inboard isolation valve, 2B21-F016. This was based, in part, on the as-left condition of this valve during the last RCS hydrostatic test. Due to the exponentially increasing leak rate, the licensee chose to electrically backseat the MOV. On January 25, 2024, the licensee was successful in limiting the RCS unidentified leak rate back to baseline by electrically backseating the MOV using a new tool to operate the actuator remotely from motor control center (MCC).
Backseating a valve is a maintenance activity allowing the stem to contact the backseat, which can either reduce or stop packing leakage. Additionally, backseating a valve can affect the valves performance (e.g., cause damage to the valve or bind it into its backseat). Since this activity can affect the valve, NUREG-1482, revision 3, Guidelines for Inservice Testing at Nuclear Power Plants, Section 4.4.2, Post-Maintenance Testing After Stem Packing Adjustments and Backseating of Valves to Prevent Packing Leakage, provides guidance for stroking the valve stem away from the backseat after the initial backseating operation to demonstrate the valve did not become bound in the backseat. Although the licensee reviewed this guidance describing an NRC approved testing method for validating a valves performance, the licensee chose not to implement this guidance. Consequently, the licensee did not stroke the valve away from the backseat after the initial backseating operation of the valve.
The licensee performed an engineering evaluation in support of backseating the valve under engineering change request (ECR) 461530. The inspectors reviewed the evaluation and noted there were two reasons for performing this evaluation. One aspect of the evaluation was to evaluate the backseating evolution, which was to be performed for the first time with a new MOV backseat relay tool (BSRT), under Work Order (WO) 5443427-01. This tool is installed at the MCC and bypasses the open limit switch to allow the stem to contact the backseat. The second purpose of the evaluation was to perform a formal technical evaluation to assess the limitation of the backseating, the potential effects on the MOV structural capability, and the valve requirements. The licensee evaluated the following three criteria:
increased stroke time, thrust and torque loads applied during backseating of the valve compared to valve/actuator structural capability, and post-maintenance requirements/testing/evaluations. From the review, the inspectors noted the structural capabilities of the valve and actuator were calculated to be acceptable within design limits.
Regarding the licensees evaluation on PMT requirements, MA-AA-716-012, Post Maintenance Testing, (PMT) revision 28, contains an MOV PMT test matrix which lists common maintenance activities, such as valve replacement, and assigns pre-established PMT verification(s) to be performed. Although neither manual nor electrical backseating were included in the test matrix, the licensee considered the listed maintenance activity of control circuit disconnect and reconnect to be applicable since this would occur during the MOV BSRT installation. Therefore, the listed test verifications included rotation and logic check, control room functional stroke, and an IST operability stroke time test. However, attachment 2 of the PMT procedure, Waiver Requirement Guidance, allowed engineering to waive a PMT provided a written justification was completed. The licensee also provided guidance for engineering to waive the PMT under procedure ER-AA-302-1006, revision 21, Motor-Operated Valve Maintenance Testing Guidelines, under step 4.2, Guidelines for Exempting Certain Motor-Operated Valve (MOV) Post-Maintenance Diagnostic (PMDT)
Recommendations from MA-AA-716-012. The written justification for waiving the referenced test verification was documented in the engineering evaluation. Therefore, no testing was performed prior to returning the valve to service after the valve was electrically backseated.
LaSalle County Station (LSCS) Inservice Testing (IST) Program Plan - 4th Interval, Revision 0, states the Code of Record for the Fourth 10-Year IST Program interval is the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM) Code, 2004 Edition through 2006 Addenda. The IST requirements apply, in part, to valves required to perform a specific function in shutting down the reactor to the safe shutdown condition, in maintaining the safe shutdown condition, or in mitigating the consequences of an accident.
The IST Program Plan - 4th Interval incorporates ASME Code Case OMN-1, 2006 Addenda, through the alternative granted by the Nuclear Regulatory Commission (NRC) in valve relief request - RV-01, Utilization of ASME Code Case OMN-1. With this alternative granted by the NRC for this IST Program interval, the inspectors noted primary containment isolation MOV 2B21-F016 was subject to ASME Code Case OMN-1, Alternative Rules for Preservice and Inservice Testing of Active Electric Motor-Operated Valve Assemblies in Light-Water Reactor Power Plants.
The IST Program Plan - 4th Interval states 2B21-F016 is an ASME Class 1, Category A, normally open, motor-operated, active valve with a safety function in the closed position. The ASME OM-2006 Code Case OMN-1, paragraph 3.4, states, in part, When an MOV or its control system is replaced, repaired, or undergoes maintenance that could affect the valves performance, new inservice test values shall be determined, or the previously established inservice test values shall be confirmed before the MOV is returned to service... This testing is intended to demonstrate that performance parameters, which could have been affected by the replacement, repair, or maintenance, are within acceptable limits. The Owners program shall define the level of testing required after replacement, repair, or maintenance...
Since both the valve and its control system underwent maintenance that could affect the MOVs performance, the inspectors determined the valve was required to be tested in accordance with Paragraph 3.4 of Code Case OMN-1. The Code Case does not provide the allowance to perform an evaluation in lieu of the required inservice testing. Although this does not conform to the requirements, the licensees evaluation and work results provided reasonable assurance the structural integrity of the MOV was not exceeded. However, not performing any testing after maintenance prior to returning the valve to service did not maintain the requisite level of assurance for the valve. Considering operating experiences with backseating, the number of assumptions embedded in the evaluation, and the use of the new MOV BSRT, the inspectors noted there were several uncertainties associated with the engineering evaluation and its use to restore operability. Based on the inspectors review of the IST plan and procedures for the valve, the inspectors determined the licensee failed to ensure the testing required after maintenance under WO 5443427-01 was performed in accordance with Code Case OMN-1. In addition, the licensee did not request relief from the code via an ASME Code relief request to the NRC which, if approved, would have allowed the valve to be returned to service without performing the required testing.
Corrective Actions: The licensee entered this issue into their corrective action program. The licensee is evaluating the technical and regulatory requirements for resolution and alignment.
Corrective Action References: AR 4754319, NRC ID ASME OM Code Potential Finding/Violation; and AR 4753350, NRC IS Questions on Valve Backseating Activity
Performance Assessment:
Performance Deficiency: The licensees failure to perform required testing after maintenance for the primary containment isolation MOV 2B21-F016 in accordance with ASME OM Code-2004, 2006 Addenda, Code Case OMN-1, Paragraph 3.4, was a violation of 10 CFR 50.55a(f)(4)(ii) and a performance deficiency. Specifically, the licensee failed to perform required testing on the MOV prior to returning the valve to service after electrically backseating the valve, which was maintenance that could affect the valves performance.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the SSC and Barrier 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. The inspectors determined the finding was associated with the SSC and Barrier Performance objective of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, by not performing the required testing, the licensee did not maintain the requisite level of assurance of the valves reliability of performing its intended function after performing maintenance that could affect the valves performance.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined the finding was of very low safety significance (Green) because they answered No to all exhibit 3, Barrier Integrity Screening Questions, section C, Reactor Containment, screening questions.
Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, when presented with an emergent and exponentially increasing RCS unidentified leakage, the licensee determined testing after maintenance was not required prior to returning the valve to service through an engineering evaluation. This was the licensees first time performing this evolution and using the new MOV BSRT, and the licensee did not adequately challenge, understand, and manage the activity affecting quality to ensure regulatory requirements were met.
Enforcement:
Violation: Title 10 CFR 50.55a(f)(4)(ii), requires, in part, Inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, conducted during successive 120-month intervals must comply with the requirements of the latest edition and addenda of the ASME OM Code incorporated by reference in paragraph (a)(1)(iv) of this section 18 months before the start of the 120-month interval (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.192 as incorporated by reference in paragraph (a)(3)(iii) of this section).
LaSalle County Station IST Program Plan - 4th Interval, Revision 0, establishes the Code of Record for the Fourth 10-Year IST Program Interval (October 12, 2017 - October 11, 2027)as the ASME OM Code, 2004 Edition through 2006 Addenda, as incorporated by reference in 10 CFR 50.55a. LSCS submitted a Valve Relief Request RV-01 to implement the optional ASME Code Case OMN-1 in their Fourth 10-Year IST Program Plan.
ASME OM Code-2006, Code Case OMN-1, Paragraph 3.4, Effect of MOV Replacement, Repair, or Maintenance, states, in part, When an MOV or its control system is replaced, repaired, or undergoes maintenance that could affect the valves performance, new inservice test values shall be determined, or the previously established inservice test values shall be confirmed before the MOV is returned to service.
Contrary to the above, on January 25, 2024, the licensees inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, did not comply with the requirements of the 2004 Edition through the 2006 Addenda of the ASME OM Code as incorporated by reference in 10 CFR 50.55a for the current 10-Year IST program interval at LaSalle County Station effective October 17, 2017. Specifically, the licensee failed to perform any testing on primary containment isolation MOV 2B21-F016, a valve within the scope of the ASME OM Code and Addenda, before returning the valve to service after performing a maintenance activity that could affect the valves performance. The maintenance activity involved disconnecting and reconnecting the MOVs circuitry as part of installing and removing an MOV backseat relay tool used to electrically backseat the valve.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Use Calibrated Measuring and Test Equipment to Electrically Backseat a Safety Related Motor-Operated Valve Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000374/2024001-03 Open/Closed
[H.11] -
Challenge the Unknown 71111.24 The inspectors identified a finding of very low safety significance (Green) and a NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XII, Control of Measuring and Test Equipment, for the licensees failure to assure a tool used in activities affecting quality was properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits. Specifically, the licensee failed to apply quality assurance requirements to the MOV BSRT when using the tool to electrically backseat a Unit 2 safety-related valve.
Description:
On January 25, 2024, the licensee used a tool, the MOV BSRT, to electrically backseat the B MSIV drain header inboard isolation valve 2B21-F016, a safety-related containment isolation MOV. The valve was backseated to address a suspected packing leak, which was contributing to an increase of Unit 2 RCS unidentified leakage. Although backseating is a known method available to mitigate valve packing leaks, backseating is not commonly performed. Furthermore, electrically backseating is an even more uncommon evolution. In both manual and electrical backseating of valves, operating experience has shown backseated valves can be either damaged or become bound into their backseats. The NRC has provided guidance on backseating under section 4.4.2, Post-Maintenance Testing After Stem Packing Adjustments and Backseating of Valves to Prevent Packing Leakage, of NUREG-1482, revision 3, Guidelines for Inservice Testing at Nuclear Power Plants.
The inspectors performed a review of the MOV BSRT used to accomplish this activity. Engineering evaluation documented under ECR 461530 was reviewed and found to incorporate guidance of draft procedure MA-AA-723-304, Electrical Backseating Motor-Operated Valves Remotely from a Motor-Control Center - Current Cut-Off Method, for using this tool. From the evaluations conclusion, the inspectors noted this was the first use of the MOV BSRT on a currently installed valve at the site. The guidance from the referenced draft procedure was also incorporated in WO 5443427-01, which was implemented to perform the electrical backseating of the valve with the MOV BSRT. The MOV BSRT was installed remotely at the MOVs MCC on the open control circuit in parallel with the Limitorque actuator open limit and torque switch. This allowed the tool to bypass the MOV open limit settings to stroke the valve in the open direction. Normally, during an open stroke of the valve, the valves actuator would stop the valve travel near the fully open position when the open limit switch trips before the stem contacts the backseat. The MOV BSRT uses a specified threshold percentage of minimum recorded running current to stop the valve travel after the valve stem contacts the backseat.
The inspectors questioned the quality of the MOV BSRT used for this maintenance evolution as this was an activity affecting quality. The licensee stated the MOV BSRT was not considered measurement and testing equipment (M&TE) per procedure MA-AA-716-040, revision 16, Control of Portable Measurement and Test Equipment Program. The licensee also stated, The backseat relay tool monitors current and then performs its function in terms of percent. Since it gives functions in terms of percentages and not absolute terms, no calibration for current reading is necessary.
The inspectors reviewed the MOV BSRT users manual, TM201602, revisions 9 and 13. In both revisions, section 9, Maintenance, states, in part, Calibration of the device may be performed but is not required. Trip setpoints are specified as a percentage of current. As such, the actual current does not matter to meet the intent. Current readings are provided for information. Refer to the calibration procedure available at the link below. The licensee determined the Quality Assurance (QA) required calibration and verification of the tool measured current would not be performed based on the users manual. Through discussions with the licensee, the inspectors noted that the users manual step 4.9 of revision 13 the licensee later referenced was not found in the evaluation, which referenced revision 9.
Revision 13, step 4.9, states, Since all setpoints are expressed as percentages, accurate calibration of the relay or probes is not relevant to trip functioning. The inspectors also reviewed the MOV BSRT functional test and calibration procedure, TP201602-03, revision 1, and found the stated purpose was to provide a means of bench verifying the relay responds correctly to current inputs. The test also verifies that each phase current input is operational as well as the displayed value of current. Additionally, the procedure states, Since the relay operates on relative current readings to detect increased motor load as the valve reaches the backseat, calibration of current reading is not required for proper functioning of the device.
Therefore, although the current reading displayed is not required for proper functioning of the tool, it is still necessary to test and calibrate the MOV BSRT to ensure both the relays operate correctly and the phase current inputs are operational.
When Constellation procured the MOV BSRT in 2019, the tool was classified as the following:
For General Use Only, Not for Qualitative or Quantitative Measurements, and Indication Only. After purchasing from a commercial vendor, the MOV BSRT was entered into the companys M&TE log for tool traceability. However, the licensee did not establish QA related controls for the tool.
The inspectors reviewed the following quality assurance requirement. Title 10 CFR Part 50, Appendix B, Criterion XII, states: Measures shall be established to assure that tools, gages, instruments, and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits. Since the tool was used in activities affecting quality, the inspectors determined the licensee failed to classify the MOV BSRT as a tool to be controlled and calibrated in accordance with 10 CFR 50, Appendix B, Criterion XII, and their Quality Assurance Topical Report (QATR) section 12, Control of Measuring and Test Equipment.
Therefore, the licensee failed to established measures to assure that the MOV BSRT, which was used in an activity affecting quality, was properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits.
Corrective Actions: The licensee has entered the inspectors concern into their corrective action program. The licensee developed and performed just-in-time training for operating the MOV BSRT prior to performing WO 5443427-01. Also, QA controlled MOV diagnostic testing equipment was installed at the MCC to monitor current in parallel with the MOV BSRT current monitoring probes. The licensee compared the data from the QA equipment to previous diagnostic testing data and determined the tool operated as expected.
Corrective Action References: AR 4754320, NRC ID Use of Backseating Tool Potential Finding/Violation
Performance Assessment:
Performance Deficiency: The licensees failure to establish quality assurance measures for the MOV backseat relay tool used to electrically backseat a safety-related valve was contrary to 10 CFR 50, Appendix B, Criterion XII, and was a performance deficiency. Specifically, the licensee failed to assure the MOV BSRT was properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits.
Screening: The inspectors determined the performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, continued use of an uncontrolled and uncalibrated MOV BSRT has the potential to cause structural damage to a valve.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined the finding was of very low safety significance (Green) because they answered No to all exhibit 3, Barrier Integrity Screening Questions, section C, Reactor Containment, screening questions.
Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, when electrically backseating a safety-related valve, the licensee did not evaluate and manage the risk of using an unqualified tool to perform a maintenance activity affecting quality.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion XII, Control of Measuring and Test Equipment, requires, that measures shall be established to assure that tools, gauges, instruments, and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits.
Contrary to the above, on January 25, 2024, the licensee failed to assure that a tool used in activities affecting quality was properly controlled, calibrated, and adjusted to maintain accuracy within necessary limits. Specifically, the licensee failed to establish quality assurance measures for the MOV backseat relay tool used to electrically backseat a safety-related valve.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Unresolved Item (Closed)
Clarification of the Dew Point Specification for the MSA Firehawk M7 SCBA System URI 05000373,05000374/2023002-01 71124.03
Description:
The inspectors reviewed a previously identified unresolved item for a condition where the quality of breathing air used to fill self-contained breathing apparatus (SCBA) bottles did not meet all of the parameters specified by the manufacturer as specified in the user instructions manual. National Institute for Occupational Safety and Health (NIOSH) regulations and guidance state that user instructions are included as part of the NIOSH approval. Nuclear Regulatory Commission regulations require that NRC licensees use NIOSH approved equipment in their respiratory protection programs, or that they obtain approval from the USNRC to use equipment that has not been approved by NIOSH. However, it was not clear if the NIOSH approval was contingent upon the dew point guidance that applied to Grade D air, or a dew point of -65°F, two parameters of breathing air quality that conflicted with each within the user instructions.
Review of this issue was discontinued in accordance with the Very Low Safety Significance Issue Resolution (VLSSIR) process as documented in this report. No further evaluation is required.
This item is closed.
Corrective Action Reference(s): AR 4686286 Very Low Safety Significance Issue Resolution Process: Very Low Safety Significance Issue Resolution Process: Dew Point Specification for the MSA M7XT SCBA System 71124.03 This issue is a current licensing basis question and inspection effort is being discontinued in accordance with the Very Low Safety Significance Issue Resolution (VLSSIR) process. No further evaluation is required.
Description:
The NRC promulgated requirements for the use of respiratory protection and controls to restrict internal exposure in Subpart H to 10 CFR 20 Standards for Protection Against Radiation. Within this regulation are requirements when a licensee assigns or permits the use of respiratory protection equipment to limit the intake of radioactive material. Some anticipated uses of respiratory protective equipment to reduce the intake of radioactive material include repair of highly contaminated equipment, decontamination of large surface areas, and responding to an accident or a fire involving radioactive contamination. The respiratory protection equipment with highest protection factor is the SCBA. These are used for responding to an accident or a fire involving radioactive contamination. The SCBA unit is a device that includes a face mask connected to a bottle of compressed air carried on the back of the user. This is also known as an atmosphere-supplying respirator, as the only air available to the user is from the bottle of compressed air.
Atmosphere-supplying respirators, such as SCBAs, must be supplied with respirable air of Grade D quality or better as defined by the Compressed Gas Association in publication G-7.1, Commodity Specification for Air, 1997 and included in the regulations of the Occupational Safety and Health Administration (29 CFR 1910.134(i)(1)(ii)(A) through (E)).
Grade D quality air criteria include
- (1) oxygen content (v/v) of 19.5-23.5%
- (2) hydrocarbon (condensed) content of 5 milligrams per cubic meter of air or less
- (3) carbon monoxide (CO) content of 10 ppm or less
- (4) carbon dioxide content of 1,000 ppm or less
- (5) lack of noticeable odor Compressed Gas Association in publication G-7.1, Commodity Specification for Air, 1997 Table 1 - Directory of Limiting Characteristics, also includes maximum dew point or moisture content for compressed air used as breathing air. The value listed in the table for Grade D air is blank but covered by a footnote from the dew point parameter. Specifically, this states The water content of compressed air required for any particular quality verification level may vary with the intended use from saturated to very dry. For breathing air used in conjunction with a self-contained breathing apparatus in extreme cold where moisture can condense and freeze causing the breathing apparatus to malfunction, a dew point not to exceed -65°F (24 ppm v/v) or 10 degrees Fahrenheit lower than the coldest temperature expected in the area is required. If a specific water limit is required, it should be specified as a limiting concentration in ppm (v/v) or dew point...
The inspectors have observed test results for breathing air quality consistently achieved results with moisture content between 63 ppm (v/v) and 24 ppm (or dew point between -50°F and -65°F).
The inspectors reviewed the operation and instructions manual published by the SCBA manufacturer to identify limitations or evaluations that might assist with the evaluating the apparent failure to ensure the SCBA will remain functional if used in extreme cold where moisture can condense and freeze causing the breathing apparatus to malfunction. The inspectors identified inconsistencies as it pertains to dew point specifications for SCBA.
Specifically, within the same user instruction documentation, in one section of the instructions include a more conservative dew point might be prescribed when compared to the dew point specified in another section. The least stringent dew point inspectors observed corresponds to Grade D air (i.e., -50°F or 10°F) less than coldest expected ambient temp), whereas the more conservative dew point corresponds to that of Grade L air (i.e., -65°F). Additionally, the manual includes a special or critical user instruction that states the equipment is approved for use at temperatures above -25°F.
The licensee has revised air quality testing procedure parameters. Since these changes were implemented, the licensee has demonstrated the breathing air used to fill bottles used for SCBA consistently satisfies the more stringent standard (-65°F).
Licensing Basis: The requirements for use of respiratory protection equipment to limit the intake of radioactive material are established in 10 CFR 20.1703. Specifically, § 20.1703 states -
- (a) The licensee shall use only respiratory protection equipment that is tested and certified by the National Institute for Occupational Safety and Health (NIOSH) except as otherwise noted in this part.
User instructions are part of the NIOSH certification; therefore, the inconsistency in the user instructions introduces a situation where the equipment cannot be used per its instructions; presumably leading to a violation of the NIOSH certification and our requirements.
Significance: The inspectors determined the issue was of very low safety significance because the less stringent dew point specification of Grade D air (-50°F) was sufficient for the environment (ambient temperatures above -25°F) in which the equipment was used or would be used. Additionally, some of this equipment has been in service for many years without issue.
For the purpose of the VLSSIR process, the inspectors screened the issue of concern through IMC 0609, Appendix C and determined the issue of concern would likely be Green had a performance deficiency been identified. Specifically, it would not have been an as-low-as-reasonably-achievable planning issue, there would not have been overexposures, nor substantial potential for overexposures, and the licensees ability to assess dose would not be compromised. Therefore, the condition represents an issue of very low safety significance that does not warrant additional review.
Technical Assistance Request: The inspectors did not enter either the TIA or technical assistance request process. However, the inspectors contacted the cognizant branch in the Office of Nuclear Reactor Regulation (NRR). Attempts to resolve whether the inconsistency in the user instructions introduced a situation where the equipment cannot be used per its instructions or invalidated the NIOSH certification were inconclusive. Consequently, the inspectors could not determine whether the respiratory protection equipment was used as certified by the NIOSH and required by 10 CFR 20.1703(a).
Corrective Action Reference: AR 4686286 Failure to Promptly Correct Degraded Pressure Switches in the Unit 1 and Unit 2 Main Steam Line High Flow Isolation Logic System Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000373,05000374/2024001-04 Open/Closed None (NPP)71152A The inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly correct a condition adverse to quality associated with the Unit 1 and Unit 2 main steam line (MSL) isolation logic system. Specifically, the licensee identified in a 2008 equipment apparent cause evaluation that pressure switches installed in the Unit 1 and Unit 2 MSL high-flow isolation logic trip systems were susceptible to multiple failure modes and had been exposed to peak inductive currents during frequent calibration activities that may have damaged switch contacts. A subsequent corrective action replaced half of the impacted switches while the other half continued to be replaced on an as-needed basis.
Description:
On January 9, 2024, LaSalle County Station (LSCS) Unit 2 received a main steam line isolation valve (MSIV) half-isolation signal from the A trip system during a scheduled MSL high-flow isolation calibration activity. Troubleshooting performed by the licensee subsequently identified a failed pressure switch in the MSL high-flow isolation logic that they believe caused the spurious half-isolation signal.
There are four differential pressure switches connected to each of the four MSLs in the MSL high-flow isolation logic system, for a total of 16 differential pressure switches per unit. Each of the four pressure switches connected to a steam line input to a trip channel and there are two trip channels in each trip system. A half-isolation signal with no associated MSIV closure occurs when the relays in a single trip system drop out. Currently, the licensee uses pressure switches manufactured by Static-O-Ring (SOR). Two SOR models are installed in the trip channels including the 102 series that was installed starting in the 1980s. The licensee began to replace the 102 series with the 131 series on an as-needed basis in 2005 due to internal diaphragm failures, contact quality issues, and setpoint drift experienced with the 102 series.
The resident inspectors evaluated the maintenance history of the failed pressure switch and concluded that it was a 102-series SOR switch and was 22 years old at the time of the spurious Unit 2 half isolation. Further, the resident inspectors reviewed an equipment apparent cause evaluation (EACE) from 2008 for a similar spurious MSIV half-isolation signal that occurred on Unit 1 in the B trip system. The EACE notes that the half-isolation signal was caused by a failed pressure switch. The EACE also notes that the faulty pressure switch removed after the event presented with contacts that were severely arc damaged. It suggests that observed damage was caused by peak inductive currents generated during the calibration activity. In 2008, LaSalle technical specifications only required this calibration activity every 2 years, though the licensee was performing it every 92 days to address setpoint drift exhibited by 102 series switches as noted above. Thus, the EACE suggests that the observed damage to switch contacts was caused by the accelerated frequency of the pressure switch calibration activity.
The cited apparent cause listed in the 2008 EACE was the design of the MSL High Flow Switchis susceptible to multiple failure modes. The basis associated with a subsequent causal factor further notes that shortly after the 2008 half-isolation, only 8 of the 32 (16 per unit) MSL pressure switches had been upgraded to 131 series and reflects that if the replace as they fail philosophy were not employed, this event could have been avoided.
Corrective actions listed in the EACE included replacement of the failed defective flow switch and implementation of a replacement schedule so that at least one string in each trip system will be replaced in approximately 6 months. An action item to develop a test box was further listed in the EACE that could be installed during the calibration activity and suppress the inductive current experienced by the switch contacts.
The residents reviewed maintenance records associated with the MSL isolation high-flow pressure switches and determined that all Unit 1 and Unit 2 pressure switches in the B1 and B2 trip channels were replaced by May 2009 in response to the corrective actions noted above. They reviewed calibration procedures and noted that the test box identified by another corrective action was developed and implemented in May 2011 via test procedure LIS-MS-102/202, Unit 1/Unit 2 Main Steam Line High Flow MSIV Isolation Calibration.
The resident inspectors note however, that at the time of the January 2024 Unit 2 MSIV half-isolation, only 8 of the 16 Unit 1 and Unit 2 pressure switches on the A1 and A2 trip channels had been upgraded to the 131-series model. In fact, switches in these trip channels have a mean in-service age of 28 years. The resident inspectors have not identified a corrective action assigned to the pressure switches in these trip channels, even those the licensee concluded in the 2008 EACE that the half-isolation was caused by the multiple failure modes of the SOR model 102 pressure switches and frequent calibration activities without surge protection most likely induced degradation across the pressure switch contacts.
Instead, these switches have been replaced as needed based on quarterly calibration testing.
The inspectors also reviewed several other maintenance issues associated with the MSL high-flow pressure switches that occurred between 2008 and 2024. In 2014, the licensee identified that both the 102 and 131-series SOR switches were exhibiting an increasing trend in diaphragm failures. Another EACE documented in 2016 logged a series of calibration issues seen in both 102 and 131-series SOR switches. An action item resulting from that EACE produced a project to replace SOR-manufactured switches with Rosemount Transmitters and trip units. Although this project was originally planned to be implemented in approximately 2018, the upgrade has been delayed until 2028. Currently, the licensee maintains a quarterly calibration frequency and is replacing pressure switches with the SOR 131-series model on an as-needed basis.
Corrective Actions: In response to the January 2024 half-isolation signal, the licensee replaced the bad SOR 102-series switch with a SOR 131-series switch. They also performed a failure analysis on the bad switch and determined that the half-isolation signal occurring in 2024 was most likely caused by a poor solder connection at the micro-switch.
In response to the inspectors observations, the licensee wrote corrective action AR 4761815 to evaluate Unit 1 and Unit 2 MSIV isolation pressure switches that have not been replaced since May 2011 and generate work requests as needed.
Corrective Action References: ARs 4727857, 4730751, 844283, 2607807, and 4761815
Performance Assessment:
Performance Deficiency: The inspectors determined that the licensees failure to promptly replace SOR pressure switches in the A1 and A2 MSL high-flow trip channels that were exposed to potential degradation across contacts is a performance deficiency. Specifically, a 2008 EACE evaluated a spurious MSIV half-isolation similar to the isolation that occurred in January 2024 and cited the known multiple failure modes associated with the switches as an apparent cause of the 2008 half-isolation. It also identified that the switch contacts had been exposed on multiple occasions to peak inductive currents, potentially causing degradation to those contacts. A corrective action to replace half the impacted switches was identified and implemented at the time. The remainder of the switches have been replaced on an as-needed schedule. To date, seven pressure switches on the Unit 1 and Unit 2 MSLs have not been replaced.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failure to replace SOR pressure switches in the A1 and A2 MSL high-flow trip channels after identifying that the switches were susceptible to multiple failure modes and had been exposed to peak inductive currents increased the potential for a spurious MSL isolation and reactor scram. Unit 1 is currently operating with five pressure switches which were subjected to the noted degradation mechanism and Unit 2 is currently operating with two such switches.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The finding screened as Green, or very low safety significance, because the inspectors answered No to Exhibit 1, Section B questions. To date, no full MSIV isolation/reactor trip has been attributed to faulty pressure switches at LaSalle.
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: 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, states that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
Contrary to the above, as of April 17, 2024, the licensee failed to establish measures to assure that conditions adverse to quality are promptly corrected. Specifically, the licensee failed to correct pressure switches installed in the Unit 1 and Unit 2 MSL high-flow isolation logic trip systems that are susceptible to multiple failure modes and have been exposed to peak inductive currents during frequent calibration activities that may have damaged the switch contacts. The licensee replaced all switches in the B1 and B2 trip channels while switches in the A1 and A2 trip systems have been replaced on an as-needed basis.
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 17, 2024, the inspectors presented the integrated inspection results to Christopher Smith, Plant Manager, and other members of the licensee staff.
- On March 1, 2024, the inspectors presented the radiation protection inspection results to John VanFleet, Site Vice President, and other members of the licensee staff.
- On March 1, 2024, the inspectors presented the inservice inspection results to John VanFleet, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
M-96, Sheet 1
P&ID Residual Heat Removal System (RHRS)
Drawings
M-96, Sheet 4
P&ID Residual Heat Removal System (RHRS)
AH
Procedures
LOS-RH-Q1
RHR (LPCI) and RHR Service Water Pump and
Valve Inservice Test for Modes 1, 2, 3, 4, and 5
101
FZ 2K
Rx Bldg. 687'-0" to 768'-0" Elev. U1 Steam Tunnel
Fire Plans
FZ 3I4
Rx Bldg. 673'-4" Elev. U2 LPCS/RCIC Pump Cubicle
Engineering
Changes
Evaluate Plugging Floor Drains to Support Maintenance in
the Replacing of the U1 L1R20 250V Batteries (1DC01E)
2/08/2024
Ultrasonic Examination of RPV Recirculation Outlet Nozzle
IR 1-NIR-10
2/24/2024
Ultrasonic Examination of RPV Core Spray Nozzle
IR 1-NIR-16
2/25/2024
Ultrasonic Examination of RPV Bottom Head Meridional
Weld GEL-1006-DB
2/27/2024
Ultrasonic Examination of RPV Bottom Head Meridional
Weld GEL-1006-DC
2/27/2024
Ultrasonic Examination of RPV Bottom Head Meridional
Weld GEL-1006-DD
2/27/2024
Visual Examination of MS Restraint MS01-1352X
2/21/2024
NDE Reports
Visual Examination of RHR Constant Support RH40-1004C
2/23/2024
GEH-PDI-UT-1
PDI Generic Procedure for the Ultrasonic Examination of
Ferritic Welds
2.1
GEH-PDI-UT-2
PDI Generic Procedure for the Ultrasonic Examination of
Austenitic Pipe Welds
GEH-UT-300
Procedure for the Manual Ultrasonic Examination of Reactor
Procedures
GEH-UT-311
Procedure for Manual Ultrasonic Examination of Nozzle
Inner Radius, Bore, and Selected Nozzle to Vessel Regions
Work Orders
Unit 1 RHR HX Weld 1RH-HX1B-9A Repair
03/04/2022
LGP-1-1
Normal Unit Startup
134
NF-AB-720-F-1
L1C21 Startup Sequence
Procedures
L1C21 BOC Startup ReMA
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
1B RR Seal Pressure and Temperature Fluctuations
1A RR Seal Pressure and Temperature Fluctuations
LAS Named in 10CFR Part 21 Notification from
Valcor Eng Co.
2/27/2023
Metal Shavings Discovered during 2B DG Top Deck
Inspection
01/23/2024
Corrective Action
Documents
GGJ319 GNF Water Rod Inspection Not Complete
2/21/2004
Work Orders
WO 8003976-fa
Failure Analysis of Woodward EGR Actuator
Work Orders
L1R20 Water Rod Inspections
2/22/2024
Seismic Qualification of Velan 3" Class 900 MO Gate Valves
for Use in Applications 1(2)B21-F016, 19
Calculations
LAS-2B21-F016
MIDACALC MOV Datasheet
1AP04E-13 Will Not Close - 1A RR LFMG BKR 1A
2/21/2024
Corrective Action
Documents
HYDRO Test Issues
03/04/2024
NRC ID Questions on Valve Backseating Activity
2/26/2024
NRC ID ASME OM Code Potential Finding/Violation
2/29/2024
NRC ID Use of Backseating Tool Potential Finding/Violation
2/29/2024
NRC-Identified IR for WO 5403063
03/22/2024
Corrective Action
Documents
Resulting from
Inspection
Work Order 05443427-01 Amendment
03/27/2024
Engineering
Changes
Evaluate Electrically Backseating 2B21-F016 Remotely from
Motor Control Center Using MOV Backseat Relay Tool
IST-LAS-BDOC-
V-14
IST Basis Document for 2B21-F016
03/01/2019
IST-LAS-PLAN
Inservice Testing Program Plan Fourth Ten-Year Interval
Quality Assurance Topical Report
TM201602-AC
MOV Backseat Relay for AC Motors Users Manual
TM201602-AC
MOV Backseat Relay for AC Motors Users Manual
Miscellaneous
TP201602-03
MOV Backseat Relay Model 201602-AC Functional Test and
Calibration
LIS-MS-202
Unit 2 Main Steam Line High Flow MSIV Isolation Calibration 27
LOS-DG-109
Unit 1 Integrated Division 1 Response Time Surveillance
LOS-NB-R1
U1 Reactor Vessel Leakage Test
Procedures
Post Maintenance Testing
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Control of Portable Measurement and Test Equipment
Program
Electrical Backseating Motor-Operated Valves Remotely
from a Motor Control Center - Current Cut-Off Method
IST Comprehensive Pump Test for 2E22-C002
03/18/2024
Div 1 Integrated Divisional Response Time Test
2/22/2024
ASME XI ISI of Class I Components and Associated Piping
03/05/2024
IM LIS-MS-202 U2 MSL High Flow MSIV Isolation Cal
01/11/2024
WO Needed to Perform LEP-DC-01 for 1DC 14E Cell 56
01/29/2024
LOS-SC-Q1 U2 B SBLC Pump Quarterly
03/15/2024
U2 Drywell Leakage Has Increased
01/25/2024
Work Orders
2A Diesel Generator Idle Start
03/15/2024
Work Orders
1A DG Fast Start
03/29/2024
LA-01-24-00502
Refueling Outage: Drywell Radiation Protection Dept.
Activities
LA-01-24-00513
L1R20 Control Rod Drive (CRD)/Undervessel Activities
ALARA Plans
LA-01-24-00601
L2R20 RB RWCU System Maintenance Activities
Corrective Action
Documents
APS: Unexpected Dose (WO 5236584-04)
2/20/2024
NRC Observations of Radiological Activities
2/28/2024
Corrective Action
Documents
Resulting from
Inspection
NRC Observations
2/29/2024
Miscellaneous
LA-01-24-00601
TEDE ALARA Evaluation Screening Worksheet for
L1R20 RB RWCU System Maintenance Activities 500k G/A,
dpm alpha
01/31/2024
Determination of Alpha Levels and Monitoring
Refueling Outage DW Under Vessel CRD Preps/Exchange
Protective Clothing Matrix
Procedures
Controls for Very High Radiation Areas
LA-01-24-00903
L1R20 RFF Cavity Platform IVVI and Associated Activities
04/19/2023
Radiation Work
Permits (RWPs)
LA-010-24-00601
L1R20 RB Reactor Water Clean Up System Maintenance
Shipping Records
Shipment LM24-
Control Rod Drive Boxes
2/29/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
003
RM - U1 Chemistry Sampling Indicates Potential Fuel Defect
2/28/2022
71151
Corrective Action
Documents
U1 Tertiary Oil Addition Troubleshooting
03/22/2023
EACE for Critical Component Failure of 2E31-N011D
01/05/2016
Spurious U2 A1 MSIV Isol Trip
01/09/2024
UMCRA: 2H13-P601-F504 CHAN A1/A2 MSIV ISOL Trip
01/12/2024
Corrective Action
Documents
AR 844283844283Unexpected Group 1 MSIV Half-Isolation Signal
11/12/2008
Corrective Action
Documents
Resulting from
Inspection
NRC ID - MSL Flow Switch Degradation Not Addressed
03/28/2024
Drawings
Schematic Diagram Primary Containment & Reactor Vessel
Isolation System PC (B21H) Part 2
Engineering
Changes
SOR Model Replacement/Alternative Solution
03/16/2010
Miscellaneous
Licensee Event
Report
05000374/2023-
003-00
Automatic Actuation of Reactor Protection System (RPS)
05/02/2023