IR 05000254/2024001
| ML24134A201 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 05/14/2024 |
| From: | Robert Ruiz NRC/RGN-III/DORS/RPB1 |
| To: | Rhoades D Constellation Energy Generation |
| References | |
| IR 2024001 | |
| Download: ML24134A201 (1) | |
Text
SUBJECT:
QUAD CITIES NUCLEAR POWER STATION - INTEGRATED INSPECTION REPORT 05000254/2024001 AND 05000265/2024001
Dear David Rhoades:
On March 31, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Quad Cities Nuclear Power Station. On April 23, 2024, the NRC inspectors discussed the results of this inspection with Doug Hild, Site Vice President, and other members of your staff.
The results of this inspection are documented in the enclosed report.
Three findings of very low safety significance (Green) are documented in this report. Three of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement; and the NRC Resident Inspector at Quad Cities Nuclear Power Station.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; and the NRC Resident Inspector at Quad Cities Nuclear Power Station.
May 14, 2024 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. 05000254 and 05000265 License Nos. DPR-29 and DPR-30
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000254 and 05000265
License Numbers:
Report Numbers:
05000254/2024001 and 05000265/2024001
Enterprise Identifier:
I-2024-001-0079
Licensee:
Constellation Nuclear
Facility:
Quad Cities Nuclear Power Station
Location:
Cordova, IL
Inspection Dates:
January 01, 2024 to March 31, 2024
Inspectors:
J. Cassidy, Senior Health Physicist
Z. Coffman, Resident Inspector
C. Hunt, Senior Resident Inspector
T. Okamoto, Resident Inspector
J. Park, Reactor Inspector
Approved By:
Robert Ruiz, Chief
Reactor Projects Branch 1
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Quad Cities Nuclear Power 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
Inadequate Work Practices Result in Two FLEX Generators Having the Incorrect Phase Rotation Following Surveillance Testing Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000254,05000265/2024001-01 Open/Closed
[H.14] -
Conservative Bias 71111.24 A self-revealed Green finding and associated non-cited violation (NCV) of 10 CFR 50.155,
Mitigation of Beyond-Design-Basis Events, was identified for the licensees failure to maintain the capability of equipment relied upon for the sites mitigation strategies for beyond-design-basis events. Specifically, the licensee failed to restore the correct wiring configuration on two of five FLEX diesel generators following surveillance testing. This resulted in both generators being unable to provide external power as designed during an extended loss of ac power event.
Inadvertent Disconnect of the Alternate 125 Vdc Battery during Surveillance Testing of the Normal 125 Vdc Battery Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000254,05000265/2024001-02 Open/Closed
[H.14] -
Conservative Bias 71152A A self-revealed Green finding and associated non-cited violation (NCV) of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the sites failure to perform activities affecting quality in accordance with documented procedures, appropriate for the circumstances, associated with the 125 Vdc distribution system.
Specifically, during restoration of the normal 125 Vdc battery to service following surveillance testing, the site inadvertently disconnected the alternate 125 Vdc battery from the dc bus. This resulted in both batteries being unavailable to provide back-up power for the 125 Vdc distribution system on Unit 1.
Inadvertent Isolation of the Reactor Core Isolation Cooling System due to Human Performance Error During Maintenance Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000254,05000265/2024001-03 Open/Closed
[H.12] - Avoid Complacency 71152S A self-revealed Green finding and associated non-cited violation (NCV) of Technical Specification 5.4.1.a, Procedures, was identified for the licensees failure to perform procedural steps as directed by MA-QC-IM-1-13202, Unit 1 RCIC Low Reactor Pressure Isolation Calibration and Channel Functional Test. This resulted in the inadvertent isolation of the reactor core isolation cooling system making it unavailable for on-line risk.
Additional Tracking Items
Type Issue Number Title Report Section Status URI 05000254,05000265/2023002-02 Clarification of the Dew Point Specification for the MSA Firehawk M7XT SCBA System 71124.03 Closed URI 05000265/2022090-02 Potential Technical Specification Violation Associated with the Failure of the 2-0203-3B ERV 71153 Closed
PLANT STATUS
Unit 1 The unit began the inspection period at full-rated thermal power where it remained with the exception of short-term power reductions for control rod sequence exchanges, hydraulic control unit maintenance, testing, and as requested by the transmission system operator.
Unit 2 The unit began the inspection period at full-rated thermal power. On February 22, 2024, the unit began its end-of-cycle coastdown period. The unit shut down on March 18, 2024, for refueling outage Q2R27 and remained in a shutdown condition through the end of the inspection period.
For all other periods, the unit was at full-rated thermal power with the exception of short-term power reductions for control rod sequence exchanges, hydraulic control unit maintenance, testing, and as requested by the transmission system operator.
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 onsite 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 readiness for seasonal extreme weather conditions prior to the onset of seasonal below zero temperatures for the following systems on January 8, 2024:
1. intake bay traveling screens
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (4 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
(1)1B core spray on February 7, 2024
- (2) Unit 1 high-pressure coolant injection system (HPCI) on February 7, 2024 (3)2A core spray on February 20, 2024
- (4) Unit 2 HPCI on February 21, 2024
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (7 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 (FZ) 120, station blackout building on January 17, 2024
- (2) FZ 8.2.5, Unit 2 cable tunnel on January 24, 2024
- (3) FZ 8.2.7.E, Unit 2 turbine building, elevation 615'-6," north mezzanine floor on January 31, 2024
- (4) Fire risk management system walkdowns associated with a Unit 2 station blackout diesel work window on February 15, 2024
- (5) FZ 8.2.4, Unit 1 cable tunnel on February 16, 2024
- (6) FZ 8.2.8.E, Unit 2 turbine building, elevation 639', main turbine floor outside shield wall on February 29, 2024
- (7) FZ 8.2.6.E, Unit 2 turbine building, elevation 595'-0," D heater bay on March 30, 2024
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the onsite fire brigade training and performance during an announced fire drill on February 7, 2024.
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 March 19-March 26, 2024.
- (1) Ultrasonic Examination
- Main Steam Pipe Weld, 30C-S19, Cat. R-A, Item R1.20
- Reactor Pressure Vessel (RPV) Head Nozzle Weld, N6A, Cat. B-D, Item B3.90
- RPV Head Nozzle Inner Radius, N6A, Cat. B-D, Item B3.100 Magnetic Particle Examination
- Main Steam Line Weld Attachments for Support 3001B-W-102A, Cat. B-K, Item B10.20 Visual Examination
- Main Steam Line Constant Spring Support 3001B-W-102A & B, Cat. F-A, Item F1.10C Relevant Indications Accepted for Continued Service
- In-Vessel Visual Inspection (IVVI) of Core Spray Piping Weld 4P4d Welding Activities
- Repair/Replacement of Weld No. 1 and 2, Replacement of Residual Heat Removal (RHR) Low Pressure Discharge Piping Elbow under Work Order (WO) 5266486
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 shutdown for the Q2R27 refueling outage on March 18, 2024.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator requalification training on January 18, 2024.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (1 Sample)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
- (1) Action Request (AR) 4722813, U1 EDG Failed During Performance of QCOS 6600-41, on February 2, 2024
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (6 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:
- (1) Unit 1 emergency diesel generator output breaker tripped on January 11, 2024 (2)risk management for work week 01/23/2024 on January 23, 2024 (3)shutdown safety plan for Unit 2 refueling outage on March 11, 2024
- (4) Q2R27 outage risk for week 03/18/2024 on March 18, 2024 (5)online fire risk equipment walkdown during bus 23-1 outage on March 22, 2024
- (6) Q2R27 outage risk for week 03/25/2024 on March 25, 2024
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (7 Samples)
The inspectors evaluated the licensees justifications and actions associated with the following operability determinations and functionality assessments:
- (1) Operability Evaluation (OpEval) 640529 associated with the 2B residual heat removal pump flexible rubber cooling line on January 8, 2024
- (2) AR 4716141, 2-1001-47 Calculated Closure Time Greater Than Design, on February 6, 2024
- (3) AR 4753652, HPCI Switch Resistance Out of Tolerance per AT #00626169-03, on March 4, 2024
- (4) AR 4754009, U1 HPCI Water Leaks from Turbine During Operation, on March 5, 2024
- (5) AR 4577856, Unexpected Alarm 901-5 D9, CHAH A MAIN STEAM TUNNEL Hi TEMP, on March 6, 2024
- (6) AR 4756372 AO 2-1301-12 Failed to Close Within Acceptable Limits, on March 8, 2024
- (7) AR 4760505 U2 SBO A Engine Governor Oscillating Erratically, on March 23, 2024
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02)
(1 Partial)
The inspectors evaluated the following temporary or permanent modifications:
(1)
(Partial)
Unit 2, 2A fuel pool cooling water pump temporary modification to support shutdown safety risk during Q2R27 refueling outage on March 27, 2024
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Partial)
(1)
(Partial)
The inspectors evaluated the Q2R27 refueling outage activities from March 18, 2024 to March 31, 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) (3 Samples)
- (1) MOV 1-1001-18B, minimum flow valve, following planned maintenance under WO 5071957 on February 8, 2024 (2)safe shutdown makeup flowrate test, following flow indicating controller calibrations on February 22, 2024
- (3) Unit 1 HPCI PMT, following gland seal condenser hotwell level switch replacement on February 29, 2024
Surveillance Testing (IP Section 03.01) (3 Samples)
- (1) QCOS 7500-05, SBGTS Operability Test, on January 19, 2024
- (2) QCOS 2300-13, HPCI System Manual Initiation Test, on March 4, 2024
- (3) QCOS 0100-07, Feedwater Check Valve Local Leak Rate Test CK 1(2)-0220-58A/B, 1(2)-0220-62A/B, on March 25, 2024
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
- (1) QCOS 1400-01, 1A Core Spray System Flow Rate Test, on January 10, 2024
Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)
(1)local leak rate testing on main steam isolation valves during Q2R27 refueling outage on March 18, 2024
Diverse and Flexible Coping Strategies (FLEX) Testing (IP Section 03.02) (1 Sample)
- (1) AR 4749114, Troubleshooting Results from Flex Deep Well Starter, on February 8, 2024
71114.06 - Drill Evaluation
Required Emergency Preparedness Drill (1 Sample)
(1)hostile action base drill on January 30,
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 of potentially contaminated material leaving the radiologically controlled area (RCA) and workers exiting the RCA at main control point during a refueling outage (2)licensee surveys of potentially contaminated material leaving the RCA and workers exiting the RCA at track bay contractor control point during a refueling outage
Radiological Hazards Control and Work Coverage (IP Section 03.04) (3 Samples)
The inspectors evaluated the licensees control of radiological hazards for the following radiological work:
(1)main steam safety relief valve activities (2)reactor disassembly/reassembly activities (3)reactor head flange repairs High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (4 Samples)
The inspectors evaluated licensee controls of the following high radiation areas (HRAs) and very high radiation areas (VHRAs):
- (1) Unit 2 drywell
- (2) Unit 1 main steam isolation valve (MSIV) and D heater bay
- (3) Unit 1 turbine housing
- (4) Unit 1A clean pump room 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) (2 Samples)
The inspectors evaluated the configuration of the following temporary ventilation systems:
(1)portable HEPA ventilation #22 Unit 2 reactor head flange repair area (2)portable HEPA ventilation #60 Unit 2 heater bay
Use of Respiratory Protection Devices (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated the licensees use of respiratory protection devices.
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) Licensee notification (Event Notification 56606) reporting elevated tritium levels in monitoring wells associated with the radiological groundwater protection program on July 3, 2023
- (2) AR 4722018, 5197731-01 Issues Putting U1 125 Batt. On Charger, on March 9, 2024
71152S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02)
- (1) The inspectors reviewed recent events associated with the licensees maintenance work practices that might be indicative of a more significant safety issue on March 25,
INSPECTION RESULTS
Inadequate Work Practices Result in Two FLEX Generators Having the Incorrect Phase Rotation Following Surveillance Testing Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000254,05000265/2024001-01 Open/Closed
[H.14] -
Conservative Bias 71111.24 A self-revealed Green finding and associated non-cited violation (NCV) of 10 CFR 50.155, Mitigation of Beyond-Design-Basis Events, was identified for the licensees failure to maintain the capability of equipment relied upon for the sites mitigation strategies for beyond-design-basis events. Specifically, the licensee failed to restore the correct wiring configuration on two of five FLEX diesel generators following surveillance testing. This resulted in both generators being unable to provide external power as designed during an extended loss of ac power event.
Description:
On May 20, 2023, while performing QCOS 0050-05, FLEX Well Pump Test, the FLEX deep well pump would not start as expected. The licensee entered the issue into the corrective action program under issue report AR 4679489. Further troubleshooting by the licensee over the following months ultimately revealed that FLEX generators 4 and 5 had the incorrect phase rotation which prevented them from supplying electrical power as designed to support the sites mitigation strategies for beyond-design-basis external events such as an extended loss of all ac power (ELAP). The licensee documented the incorrect phase rotation under AR 4749114 on February 9, 2024, and performed a causal evaluation.
In response to NRC Order EA-12-049, Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, dated March 12, 2012, the licensee drafted RS-13-025, Overall Integration Plan in Response to March 12, 2012, Commissioning Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049).
Subsequently, the licensee developed CC-QC-118-1003, Quad Cities Power Station Final Integrated Plan Document, in accordance with NEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, which was endorsed by the NRC.
NRC Order EA-12-049 has since been codified by 10 CFR 50.155, Mitigation of beyond-design-basis events. CC-QC-118-1003 outlines the sites strategies for the mitigation of beyond-design-basis events and is intended to meet the requirements of 10 CFR 50.155.
The site FLEX strategy for supplying power to equipment required to provide core, containment, and spent fuel pool cooling, as outlined in CC-QC-118-1003, includes the deployment of mobile 500 KW diesel generators (FLEX generators) to supply external power to plant safety-related 480 Vac systems by using the in-plant seismically qualified switchgear.
Additionally, the site deploys a FLEX generator to power a seismically qualified deep well pump which provides raw water to the FLEX distribution header for use by either unit to maintain reactor pressure vessel water level when in-plant systems for reactor pressure vessel injection are no longer available.
The site has five FLEX generators total: three are stored in the seismically qualified FLEX building (N generators) and two are stored in the N+1 building that is not seismically qualified. The FLEX generators are identical in design and capacity. Any FLEX generator can be used to fulfill any function of the sites overall FLEX strategy. Per the sites strategy, one FLEX generator per unit is required to supply external power to the in-plant 480 Vac systems and one FLEX generator is required to supply power to the FLEX deep well pump during an ELAP event. The remaining two FLEX generators are administratively considered N+1 and are available as backups should issues arise in the deployment of the N FLEX generators.
The licensees causal evaluation from February 9, 2024, determined that during the set up for performing QCOS 0050-05, in May of 2023, plant personnel noted that the vendor supplied load cart did not contain the expected electrical connections as outlined in Attachment 2 of the procedure. As a result, electrical maintenance technicians had to lift electrical leads on FLEX generators 4 and 5 to connect to the vendor load cart in order to complete the surveillance as scheduled. Following the completion of surveillance testing, technicians incorrectly re-landed the leads, which inadvertently altered the configuration of the generators. As a result, both FLEX generators 4 and 5 were restored to service with the incorrect phase rotation.
The inspectors reviewed the licensees causal evaluation and the circumstances surrounding the event. The inspectors noted that there were no procedural steps in QCOS 0050-05 to lift and land leads. As a result, once the surveillance was complete, the technicians rewired the affected FLEX generators based on their knowledge of what they thought was correct.
NRC Information Notice (IN) 84-37, Use of lifted leads and jumpers during maintenance or surveillance testing, was communicated by the NRC to alert the industry of the potential for a significant degradation of safety associated with the use of lifted leads or jumpers during maintenance or surveillance testing. IN 84-37 included actions to reduce the likelihood of improperly controlled leads during maintenance. One such action was to include additional procedural checks on the system configuration during surveillance testing and maintenance.
For example, recording the initial and final positions of leads on equipment and/or adding precautionary statements to procedures regarding oversight.
Licensee procedure MA-AA-716-100, Maintenance Alteration Process, provides the administrative controls required to ensure equipment integrity when performing maintenance alterations associated with repair, replacement, or surveillance activities on plant equipment.
MA-AA-716-100 is applicable when the alternations are not controlled by other approved procedures or the work control process. The lifting and landing of electrical leads, as which was performed to support QCOS 0050-05, is under the scope of MA-AA-716-100 which directs technicians to perform such maintenance alterations in accordance section 4.2, Use of Maintenance Alterations Log, and record the activities performed on form MA-AA-716-100-F-01, Maintenance Alterations Log.
Based on the licensees casual evaluation and questions asked by the inspectors, MA-AA-716-100 was not used by the technicians performing the re-wiring of the FLEX generators to support surveillance testing because it was not procedurally directed from QCOS 0050-05, despite its applicability to the activity being performed. Rather, the technicians made the non-conservative decision to proceed with the re-wiring of the FLEX generators without a formal method to track the activity to ensure that the original configuration of the generators was properly restored once the surveillance was complete.
In addition, on April 7, 2017, during a fleet FLEX readiness review documented in AR 3995521, the licensee corporate office identified several readiness issues in the fleet regarding FLEX equipment. One such issue was the identification of the incorrect phase rotation on FLEX generators in the fleet. All sites were directed by the corporate office to physically perform the following and document the results: 1) perform a physical check, using a phase rotation meter, of the portion of the plant electrical distribution system used to power FLEX equipment, 2) perform a physical check, using phase rotation meter, of each FLEX generator that would be used to provide power, and 3) perform a physical rotation direction bump checks of any equipment that may be directly connected to the FLEX generators.
The licensee recorded the successful completion of these activities in the corrective action program on May 22, 2017.
Licensee procedure QCOP 0050-08, FLEX Electrical Restoration, provides the steps required to energize the 480 Vac distribution buses from the FLEX generators during an ELAP event. The inspectors noted that the checks performed by the licensee in response to AR 3995521 become significant because QCOP 0050-08, Step D.4, specifically states that the FLEX generators and loads have been verified for proper phase rotation during installation, therefore no further rotation checks are needed prior to operation. As such, troubleshooting the improper phase rotation of FLEX generator 4 or 5 is not covered under the procedures implementing the sites FLEX strategy because the procedures assume that the phase rotations are already correct.
Because the FLEX generators are interchangeable, the site FLEX implementation procedures do not specify which FLEX generator should be deployed to any given location. During discussions with the inspectors, the site stated that FLEX generators 4 and 5 are typically used to power the FLEX deep well pump because included on the trailer for the generator are 50-foot cables specific for use with the deep well pump starter cabinets. The cables for powering the in-plant 480 Vac busses from the FLEX generators are 100-foot cables. The inspectors determined that although the length of the cable is specific to the different components powered by the FLEX generators, there was no physical or administrative restrictions on using either FLEX generators 4 or 5 to power the in-plant 480 Vac busses as long as the correct cable was used. Therefore, the inspectors concluded that the sites FLEX strategy would have been unsuccessful if one of the two degraded FLEX generators was used to power the in-plant 480 Vac busses during an ELAP event, as was procedurally allowed.
Ultimately, the inspectors determined that the licensee failed to hold itself accountable to the sites human performance standards and maintenance program expectations during the planning, performance, and oversite of the maintenance activities associated with the affected FLEX generators. As such, the inspectors determined that the failure to restore the wiring on FLEX generators 4 and 5 to the correct configuration following surveillance testing was a performance deficiency. As a result, two of the five FLEX generators credited for use during an ELAP event were not available for approximately 8 months.
This finding was identified in connection with a review of Operating Experience Smart Sample (OpESS) 2020/01.
Corrective Actions: The licensee corrected the incorrect phase rotation on FLEX generators 4 and 5 under work order 5449678.
Corrective Action References: AR 4749114, Troubleshoot Results from Flex Deep Well Starter, AR 3995521, FLEX Readiness Fleet Review, AR 4679489, EO ID: A Flex Deep Well Pump Starter Panel Not Working
Performance Assessment:
Performance Deficiency: Incorrectly wiring FLEX generators 4 and 5 following surveillance testing was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure that FLEX generators 4 and 5 had the correct phase rotation following activities involving lifting and landing electrical leads to support surveillance testing.
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 screened the finding against the Flexible Coping Strategies (FLEX) screening questions under section E, in exhibit 2, and answered Yes to screening question E.2.
Therefore, the finding screened to needing a detailed risk evaluation.
A Region III senior reactor analyst performed a bounding risk evaluation to assess the significance of the finding. Since the finding resulted in one of the three N FLEX diesel generators having an incorrect phase rotation, each of the N FLEX diesel generators was assumed to have a 1/3 failure-to-start probability. The condition was assumed to exist for an exposure period of 8 months. The senior reactor analyst used a modified version of Standardized Plant Analysis Risk model, revision 8.82, to analyze the finding. The model modification removed credit for N+1 FLEX equipment because it was assumed insufficient time existed to deploy N+1 given a failure of the N FLEX equipment. No credit for equipment recovery was assigned.
The change in core damage frequency for the exposure period was estimated to be less than 1E-7/year. As a result, risk contributions for external events and Large Early Release Frequency were not evaluated. The dominant sequences involved weather-related loss of offsite power, failure of the emergency diesel generators and station blackout diesel generators, successful declaration of an extended loss of AC power, failure of FLEX 480 Vac generators, failure to recover offsite or onsite power AC power within battery depletion time, and failure of turbine driven injection sources resulting in core damage. The analyst determined that the significance of the finding was less than 1E-6/year primarily because the site is equipped with station blackout diesel generators with unit cross tie capability as well as engine driven FLEX pumps available to support the severe accident water addition strategy.
Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, technicians chose to proceed with the lifting and landing of leads on the affected FLEX generators without formal actions to track that the leads were re-landed in the correct configuration because the surveillance procedure in use didnt explicitly direct them to do so.
Enforcement:
Violation: 10 CFR 50.155(b)(1), states, in part, that each applicant or licensee shall develop, implement, and maintain strategies and guidelines to mitigate beyond-design-basis external events from natural phenomena that are developed assuming a loss of all ac power concurrent with a loss of normal access to the ultimate heat sink.
10 CFR 50.155(c)(1), states, in part, that the equipment relied on for the mitigation strategies and guidelines required by paragraph (b)(1) of this section must have sufficient capacity and capability to perform the functions required by paragraph (b)(1) of this section.
CC-QC-118-1003, Quad Cities Power Station Final Integrated Plan Document, outlines the sites strategy for the mitigation of beyond-design-basis events and meets the requirements of 10 CFR 50.155.
Contrary to the above, from approximately May 20, 2023, to February 9, 2024, the licensee failed to maintain equipment relied on for the sites mitigation strategies as outlined in CC-QC-118-1003. Specifically, the licensee failed to ensure that two of the five FLEX generators had the correct phase rotation following surveillance testing. The degraded condition resulted in the two affected FLEX generators being incapable of performing their design-basis function as relied on for the sites mitigation strategies.
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 M7XT SCBA System URI 05000254,05000265/2023002-02 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 Commission 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 4683815 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 self-contained breathing apparatus (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 percent
- (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 foot note 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 self-contained breathing apparatus 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 issues 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 Task Interface Agreement 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 10CFR20.1703(a).
Corrective Action Reference: AR 4683815 Inadvertent Disconnect of the Alternate 125 Vdc Battery during Surveillance Testing of the Normal 125 Vdc Battery Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000254,05000265/2024001-02 Open/Closed
[H.14] -
Conservative Bias 71152A A self-revealed Green finding and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the sites failure to perform activities affecting quality in accordance with documented procedures, appropriate for the circumstances, associated with the 125 Vdc distribution system.
Specifically, during restoration of the normal 125 Vdc battery to service following surveillance testing, the site inadvertently disconnected the alternate 125 Vdc battery from the dc bus.
This resulted in both batteries being unavailable to provide back-up power for the 125 Vdc distribution system on Unit 1.
Description:
On December 6, 2023, following surveillance testing per QCEMS 0230-02, Unit 1 125 VDC Service Test on Normal Batteries, on the normal 125 Vdc battery for Unit 1, the licensee attempted to perform QCOP 6900-25, Transfer of Unit One 125 VDC Bus Between Normal and Alternate Battery, to bring the normal battery back on the dc bus. At the time, the alternate 125 Vdc battery was connected to the bus to support surveillance testing of the normal battery and was the credited back-up power supply for the 125 Vdc distribution system. Following shift change, electrical maintenance department (EMD) technicians proceeded to lift the positive jumper connecting the alternate 125 Vdc battery to the bus and incorrectly landed the jumper onto the positive terminal of the normal 125 Vdc battery. This resulted in both batteries being disconnected from the dc bus, rendering them unavailable to provide back-up 125 Vdc power. The licensee captured the issue in the corrective action program under issue report AR 4722018 and performed a causal evaluation.
The unit 125 Vdc distribution system supplies dc control power to circuits, switchgear, the turbine system, and safety injection systems. Each unit has a normal 125 Vdc battery and an alternate 125 Vdc battery which supply electrical power to the dc distribution system whenever the battery charger, which supplies the normal source of power, fails. The normal 125 Vdc battery is capable of supplying the required power for all dc control functions that are required for control of 4160 V breakers, 480 V breakers, various control relays, and annunciators. An alternate 125 Vdc battery is provided to allow testing of the normal 125 Vdc battery while the unit remains at power. The alternate 125 Vdc battery is also available upon the inoperability of the normal 125 Vdc battery.
The licensees causal determined that EMD technicians conducting the work failed to follow procedures and validate assumptions while performing the work. Additionally, the licensee identified that the EMD technicians were not present for the pre-job brief (PJB) that discussed the steps in the procedure which placed the normal 125 Vdc battery on a charge following surveillance testing, and operators involved in the work failed to perform a high level of awareness (HLA) brief prior to commencing work.
Specifically, QCOP 6900-25 was worked partially at the beginning of day shift on December 6, 2023, in order swap bus supply power on Unit 1 from the normal 125 Vdc battery to the alternate 125 Vdc battery. Following a successful bus transfer, the same dayshift crew began performance of QCEMS 0230-02, which spanned into the afternoon shift. Dayshift and afternoon shift turned over and a PJB was performed addressing the remaining work to be performed per QCEMS 0230-02. Following the completion of QCEMS 0230-02, the afternoon shift continued to work QCOP 6900-25 to place the normal 125 Vdc battery on a charge prior to transferring the dc bus back to the normal battery. At that time, operators made the decision to omit an HLA brief prior to proceeding with the work to recharge the normal battery. Additionally, the two EMD technicians who were going to perform the work were not present during the PJB.
During execution of Step F.4.f.(1) in QCOP 6900-25 which states, Remove insulation from unattached end of long + jumper cable. The two EMD technicians disconnected the short jumper connecting the alternate 125 Vdc battery to the bus, resulting in the loss of all back-up 125 Vdc power supply.
The inspectors reviewed the licensees causal evaluation under issue report AR 4722018 and the circumstances surrounding the event. The inspectors determined that the licensee chose to not perform PJBs or HLA briefs with all of the individuals involved in restoring the normal 125 Vdc battery to service. Additionally, the licensee failed to apply general work practices for lifting and landing jumpers on safety-related equipment.
Licensee procedure, HU-AA-1211, Pre-Job Briefings, describes the performance of various PJBs to promote safe and error-free conduct of work. Specifically, an HLA brief is defined as the second highest level of pre-job briefing conducted in the plant and shall include Shift Manager participation or oversight as outlined by HU-AA-1211, section 3.2.1. However, operators involved in the work rationalized the assumption to omit an HLA brief due to the simple nature of the work. In contrast, QCOP 6900-25, section C.3, states, as a prerequisite, to conduct an HLA that includes at a minimum: 1) operations personnel involved in procedure, 2) electrical maintenance personnel involved in procedure, 3) a review of QOA 6900-02, and 4) the precautions for working on energized equipment.
In addition, a PJB was conducted without the presence of the EMD technicians involved with the work despite the work being a first-time evolution for all EMD technicians involved, as well as the first-line supervisor. The inspectors noted that HU-AA-1211 lists examples for when PJBs are recommended. Examples include using PJBs for jobs unfamiliar to the worker(s),jobs requiring coordination of two or more people or departments who will actually perform a portion of the task, and jobs with potential affects in either equipment or areas in the plant.
This includes positionable components that could be mispositioned within 2 feet of work being performed.
NRC Information Notice (IN) 84-37, Use of Lifted Leads and Jumpers During Maintenance or Surveillance Testing, provides guidance on lifting leads and jumpers on safety-related equipment. IN 84-37 specifically recommends review of procedures to ensure the instructions for surveillance and maintenance explicitly and unambiguously specify the reconnection of any lifted leads and the removal of any jumpers. Additionally, it recommends reviewing with operators and maintenance personnel specific instances of errors involving lifted leads or jumpers and the safety impact of such errors.
Overall, the inspectors determined that the licensee failed to hold itself accountable to the sites work performance standards and maintenance work practices in a manner that supported an error-free environment. The licensee rationalized the decision to omit an HLA brief, and further rationalized conducting a PJB without EMD technicians involved with the work, despite the fact it was a first-time evolution for both EMD technicians and first-line supervisor because of the perceived simple nature of the activities to be performed. As a result, the alternate 125 Vdc battery was inadvertently disconnected from the dc bus which rendered both the normal and the alternate 125 Vdc batteries unavailable. As such, the inspectors determined that the failure to perform activities to restore the 125 Vdc distribution system to a normal configuration following surveillance testing, in accordance with applicable procedures or instructions, was a performance deficiency.
Corrective Actions: Once discovered, technicians reconnected the alternate 125 Vdc battery to the dc bus.
Corrective Action References: AR 4722018, 05197731-01 Issues Putting U1 125 Batt. on Charger
Performance Assessment:
Performance Deficiency: The inspectors determined that the failure to perform activities to restore the 125 Vdc distribution system to a normal configuration following surveillance testing in accordance with the applicable procedures or instructions was a performance deficiency.
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. The inspectors determined that the failure to perform activities to restore the 125 Vdc distribution system to a normal configuration following surveillance testing in accordance with the applicable procedures or instructions was a performance deficiency.
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 screened the finding in accordance with IMC 0609, Appendix A, Exhibit 2, Section A, and answered No to all of the screening questions. Therefore, the finding screens to very low safety significance (Green).
Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the shift rationalized the decision that an HLA brief was not needed for the work to be performed due to the perceived simple nature of the activity. Additionally, a PJB was held without the EMD technicians involved in the work, despite being a first-time evolution for both EMD technicians and the first-line supervisor.
Enforcement:
Violation: Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires in part, activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
The unit 125 Vdc batteries are safety-related components associated with the 125 Vdc distribution system and applicable to 10 CFR Part 50, Appendix B. The licensee established procedure QCOP 6900-25, Transfer of Unit One 125 VDC Bus Between Normal and Alternate Battery, as the implementing procedure for this activity affecting quality.
Step F.4.f.(1) of procedure QCOP 6900-25 states, Remove insulation from unattached end of long + jumper cable.
Contrary to the above, on December 6, 2023, the licensee failed to follow Step F.4.f(1) of procedure QCOP 6900-25. Specifically, EMD technicians inadvertently disconnected the short jumper connecting the alternate 125 Vdc battery to the dc bus while it was being credited for supplying back-up power to Unit 1 dc distribution system, resulting in the loss of all back-up 125 Vdc power supply.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Observation: Elevated Tritium Levels Groundwater Monitoring Wells 71152A On July 3, 2023, the licensee made a notification to the Illinois Emergency Management Agency and Office of Homeland Security (IEMA-OHS), and subsequently to the Nuclear Regulatory Commission (Event Notification 56606), that it had detected elevated tritium levels in monitoring wells associated with the radiological groundwater protection program on site.
From July 3, 2023, to August 30, 2023, the licensee identified and repaired the two suspected causes of the elevated tritium: 1) a leaking flange from the radwaste building waste collector tank, and 2) a chimney drain piping penetration.
The inspectors monitored the licensees response to the elevated tritium in accordance with licensee procedure EN-AA-408, Radiological Groundwater Protection Program, to include remediation activities performed and corrective actions to address the suspected causes. The inspectors determined that the methodology used to quantify the amount of radioactive material leaked was adequate. The inspectors reviewed the methodology used to characterize the plume of contamination onsite, which included observations of collection and analysis of contaminated groundwater, and determined it was adequate. The inspectors determined the calculations and sampling performed by the licensee were adequate to demonstrate the contaminated groundwater did not migrate offsite via an unmonitored pathway. Finally, the inspectors determined that the criteria, methodology, and requirements for reporting leaks and spills that contain radioactive materials were consistent with the Nuclear Energy Institute Groundwater Protection Initiative and were performed in accordance with NRC requirements.
The inspectors did not identify and findings or violations in this sample.
Inadvertent Isolation of the Reactor Core Isolation Cooling System due to Human Performance Error During Maintenance Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000254,05000265/2024001-03 Open/Closed
[H.12] - Avoid Complacency 71152S A self-revealed Green finding and associated non-cited violation (NCV) of Technical Specification 5.4.1.a, Procedures, was identified for the licensees failure to perform procedural steps as directed by MA-QC-IM-1-13202, Unit 1 RCIC Low Reactor Pressure Isolation Calibration and Channel Functional Test. This resulted in the inadvertent isolation of the reactor core isolation cooling system making it unavailable for on-line risk.
Description:
On September 22, 2023, while performing MA-QC-IM-1-13202, Unit 1 RCIC Low Reactor Pressure Isolation Calibration and Channel Functional Test, maintenance technicians inadvertently isolated the reactor core isolation cooling (RCIC) system when they connected a digital multimeter across the incorrect pressure switch. Specifically, technicians isolated pressure switch 1-1360-9A per the procedure and then inadvertently connected a digital multimeter to the incorrect connection point across the 1-1360-9B pressure switch. Once the multimeter was connected across the 1-1360-9B pressure switch, with the 1-1360-9A pressure switch already isolated, full logic was achieved and the steam supply isolation valves to the RCIC system closed, making the system unavailable for operation. The licensee documented this issue in the corrective action program under issue report AR 4704434 and performed a causal evaluation.
The RCIC system is designed to operate either automatically or manually following a reactor pressure vessel isolation to provide makeup water to the reactor pressure vessel in the case of a loss of normal feedwater and maintain water level in the reactor pressure vessel above the top of the core. The RCIC system injects with a steam turbine pump driven by steam from the reactor through the RCIC steam supply line. The RCIC steam supply line low pressure isolation function is part of primary containment isolation instrumentation and is designed to isolate the RCIC steam supply line on low steam supply pressure to limit the fission product release following a design-basis accident in which there is a break in the line.
The licensees causal evaluation determined that an inadequate use of human performance tools was the primary contributor to the event. Specifically, the technicians failed to use adequate peer check and STAR [stop-think-act-review] while working through the procedure. This caused the technician performing the work to misread a step in the procedure and not be challenged by the verifying technician. The verifying technician had become complacent and did not perform as expected to challenge the action of the technician performing the step.
The inspectors reviewed the licensees causal evaluation as well as the circumstances surrounding the event. The inspectors determined that the failure to correctly perform the procedural steps of MA-QC-IM-1-13202 as directed was a performance deficiency. The inspectors noted that although the primary containment isolation instrumentation system isolated the RCIC steam supply line as designed, the inadvertent isolation took the RCIC system from being available for online risk to unavailable, and thus adversely affected the Mitigating Systems Cornerstone objective.
Corrective Actions: Once the error was corrected, operators restored the RCIC system to available by resetting and opening the primary containment isolation valves.
Corrective Action References: AR 4704434, RCIC Isolated During Performance of MA-QC-IM-1-13101
Performance Assessment:
Performance Deficiency: The inspectors determined that the failure to correctly perform the procedural steps of MA-QC-IM-1-13202 as directed was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inadvertent isolation of the RCIC system due to not correctly performing the procedural steps of MA-QC-IM-1-13202 resulted in the unplanned unavailability of the RCIC system.
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 screened the finding in accordance with IMC 0609, Appendix A, Exhibit 2, Section A, and answered No to all of the screening questions. Therefore, the finding screens to very low safety significance (Green).
Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. Specifically, technicians failed to use the appropriate human performance tools and verification practices as outlined in licensee procedure HU-AA-101, Human Performance Tools and Verifications Practices.
Enforcement:
Violation: Technical Specification 5.4.1.a, Procedures, states that written procedures shall be established, implemented, and maintained, covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978.
Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Appendix A, Section 8.b(2), states, in part, that for the reactor core isolation system, specific procedures for surveillance tests, inspections, and calibrations should be written.
Licensee procedure MA-QC-IM-1-13202, Unit 1 RCIC Low Reactor Pressure Isolation Calibration and Channel Functional Test, is a procedure that is applicable to Regulatory Guide 1.33.
Contrary to the above, on September 22, 2023, while performing MA-QC-IM-1-13202, the licensee failed to implement MA-QC-IM-1-13202, which resulted in the isolation and unplanned unavailability of the RCIC system. Specifically, maintenance technicians inadvertently isolated the RCIC system when they connected a digital multimeter across the incorrect pressure switch.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Observation: Trend Review of Maintenance Work Practices 71152S The final NRC Safety Culture Policy Statement (SCPS) that was published on June 14, 2011, provides the NRCs expectation that individuals and organizations performing regulated activities establish and maintain a healthy safety culture that recognizes the safety and security significance of their activities, and the nature and complexity of their organizations and functions. Because safety and security are the primary pillars of the NRCs regulatory mission, consideration of both safety and security issues, commensurate with their significance, is an underlying principle of the SCPS. NUREG-2165, Safety Culture Common Language, lists the traits, attributes, and examples of a healthy nuclear safety culture.
In accordance with Inspection Procedure (IP) 71152, Problem Identification and Resolution, Section 03.02, the inspectors performed a semiannual trend analysis of recent issues in the area of maintenance work practices to identify any potential trends that might indicate the existence of a more significant safety issue. The inspectors reviewed the following issues and identified deficiencies in the NUREG-2165 safety culture behavior of Avoiding Complacency (QA.4):
AR 4749154, Control Rod N-11 Scrammed in Technicians inadvertently disconnected the air line to a hydraulic control unit during corrective maintenance on a different hydraulic control unit, resulting in an unplanned control rod insertion. The issue was entered into the corrective action program and the rod was subsequently recovered following fact finding of the event.
The inspectors did not identify a finding or violation associated with this event.
AR 4761741, OSP 2-1001-5A Bonnet Bolts Over Torqued During the installation of residual heat removal valve 2-1001-5A, contractor personnel inadvertently marked a step not applicable and then proceeded to over torque the bonnet bolts to the valve. The issue was entered into the corrective action program and the valve was repair prior to being placed into service.
The inspectors did not identify a finding or violation associated with this event.
AR 4704434, RCIC Isolated During Performance of MA-QC-IM-1-13101 During performance of the Unit 1 RCIC low reactor pressure isolation calibration and functional test, technicians inadvertently isolated RCIC when they hooked up test equipment to the wrong connection. This issue resulted in a non-cited violation documented in the Inspection Results section of this report.
The inspectors reviewed the following issues and identified deficiencies in the NUREG-2165 safety culture behavior of Conservative Bias (DM.2):
AR 4722018, 05197731-01 Issues Putting U1 Batt. On Charger Following a performance test on the normal Unit 1, 125-volt direct current (Vdc) battery, technicians inadvertently disconnected both the normal and the alternate 125 Vdc battery from the safety-related dc bus. This issue resulted in a non-cited violation documented in the Inspection Results section of this report.
AR 4749114, Troubleshooting Results from Flex Deep Well Starter Following troubleshooting activities associated with the FLEX deep well pump revealed that FLEX generators 4 and 5 had the incorrect phase rotation due to electrical leads in the generators being incorrectly re-landed following surveillance testing. This issue resulted in a non-cited violation documented in the Inspection Results section of this report.
Unresolved Item (Closed)
Potential Technical Specification Violation Associated with the Failure of the 2-0203-3B ERV URI 05000265/2022090-02 71153
Description:
As documented in the August 8, 2022 NRC Integrated Inspection Report 05000254/2022090 and 05000265/2022090, the inspectors at that time had identified an unresolved item (URI)associated with the implementation of TS Limiting Condition for Operation (LCO) 3.5.1, Condition H, when 2-0203-3B may have been inoperable prior to discovery for greater than the allowed LCO completion time.
As detailed in the 2022090 inspection report, on March 21, 2022, the licensee attempted to cycle the 2-0203-3B electromatic relief valve (ERV) for as-found testing at the beginning of refueling outage Q2R26 but the valve did not stroke as expected. Later investigation by the licensee determined that the valves solenoid actuator was binding during operation, preventing the 2-0203-3B ERV from opening. As a result of the degraded condition, the 2-0203-3B ERV may have been inoperable prior to discovery and a TS LCO 3.5.1, Condition H, violation may have occurred. This issue was unresolved pending the NRCs additional independent review and evaluation to determine whether a TS LCO violation occurred.
On November 29, 2022, the NRC informed Constellation Energy Generation, LLC (CEG) of a White finding for the failure of one of the four ERVs associated with the automatic depressurization subsystem (ADS) to actuate during surveillance testing. As a result, the valve was determined to be inoperable from April 7, 2020, until March 21, 2022.
As stated in NRC Integrated Inspection Report 05000254/2022090 and 05000265/2022090, the licensee determined that the cause of the failure was due to two conditions: a warped upper guide bracket that caused increased friction in the plunger well, and incorrect installation of the internal plunger well plastic guides which decreased clearances for the plunger to move within the plunger well. The inspectors noted that workers had identified warping on the bracket during the ERV rebuild. The workers engaged engineering on the issue but did not document it in the corrective action program. Although an Engineering Change Request (ECR 445332) was issued allowing the use of shims to prevent the warping, it was not used and instead, the workers manually strengthened the bracket before reinstalling. This was contrary to the work procedure, which would have required replacement of the bracket. Additionally, the inspectors noted that the rebuild occurred in an environment noticeably cooler than the actual use conditions and that the post-maintenance testing failed to properly account for thermal expansion of the ERV components in the actual use environment. As a result, the ERV was able to pass its post-maintenance test, but later failed surveillance testing due to thermal expansion of the components causing binding.
The inspectors also noted that the work procedure did not properly incorporate the guidance in the ERV vendor guide for correct orientation of the plastic guides during installation. This deficiency was recognized by the licensee prior to the work and there was a planned revision to the rebuild procedure clarifying the installation of the guides. However, this information was not disseminated to the workers during the rebuild, who instead relied on skill of the craft to ensure alignment of the guides.
Technical specifications are part of the license issued to each power reactor by the NRC. These specifications are derived from analyses and evaluation included in the licensees safety analysis report and set forth the specific characteristics of the facility and the conditions for its operation that are required to provide adequate protection for the health and safety of the public. The Commissions Enforcement Policy states that adequate protection is presumptively assured by compliance with NRC requirements. Compliance with NRC requirements, including regulations, technical specifications, license conditions, and Orders, provides reasonable assurance to the NRC and the public that safety and security are being maintained. Any time an LCO is not met, it is a violation of the license unless the prescribed remedial actions are taken within the prescribed time frame.
The LCO for TS 3.5.1 states, in part, that the Automatic Depressurization System [ADS]
function of five relief valves shall be operable during mode 1 or in modes 2 and 3 when steam dome pressure is > 150 psig. Condition H of TS 3.5.1 defines an Allowed Outage Time (AOT) of 14 days with one ADS valve inoperable. If that AOT is not met, then per Condition I, the plant must be in mode 3, within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The NRC Enforcement Manual, Revision 11, Part II, Section 2.1.1, Actions Involving Inoperable Equipment, discusses the enforcement of apparent TS violations in detail.
Specifically, as described in Section 2.1.1(B), when a system, subsystem, train, or component is inoperable, it is appropriate to cite directly against the TS requirement for operability. Further, TS LCOs are defined as the lowest functional capability or performance levels of equipment required for safe operation of the facility, and that a violation would exist when an LCO is not met, and all necessary actions were not completed within the Allowed Outage Time (AOT). Section 2.1.1(B)(3) states Potential enforcement should be considered based on the total duration that a condition may have existed. i.e., when the time of occurrence and the extent to which the licensee should have identified the condition earlier, is readily determined. Section 2.1.1(B)(3)(d) provides additional guidance stating, If the time between the occurrence of the condition and the discovery of the condition is greater than the AOT for that condition, then the licensee should be cited for a failure to satisfy the TS LCO.
Therefore, if a system, subsystem, train, or component is inoperable for a period of time greater than its TS AOT at the time of discovery, and it was reasonable for the licensee to have discovered the inoperability at the time of occurrence, then a violation for failure to meet the AOT of the TS LCO is warranted.
In the evaluation for the URI, the NRC noted, as discussed above, that the licensee was aware of the issues with the warping of the upper guide bracket during the ERV rebuild but had failed to properly address the issue. Similarly, the licensee was aware of the need to clarify the guidance for proper alignment of the plastic guides but failed to disseminate this information to the workers. Therefore, the NRC concluded that it was reasonable for the licensee to have discovered the inoperability at the time of occurrence, (i.e., prior to returning the ERV to service on April 7, 2020). Because the issue was not identified until surveillance testing on March 21, 2022, the AOT in TS LCO 3.5.1, Condition H was exceeded. Therefore, in accordance with the guidance in Part II, Section 2.1.1 of the NRC Enforcement Manual, the NRC has concluded that a violation of TS LCO 3.5.1, Condition H, did occur.
On May 4, 2023, the NRC completed a supplemental inspection using Inspection Procedure 95001, Supplemental Inspection Response to Action Matrix Column 2 (Regulatory Response) Inputs, concluding that the licensee had adequately determined the root and contributing causes and had appropriate actions in place to address the causes identified.
As a result, the finding and associated violation was closed, and Quad Cities Unit 2 returned to Column 1 (Licensee Response) of the Action Matrix. The results of the 95001 inspection were documented in NRC Supplemental Inspection Report 05000265/2023040.
Since the associated finding was already evaluated and closed during the 95001 inspection, this additional violation will not be considered an Action Matrix input, and this URI is considered closed.
One violation was identified.
Corrective Action Reference(s): The associated CAP items for this issue were documented in the 95001 NRC Supplemental Inspection Report 05000265/2023040.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On April 23, 2024, the inspectors presented the integrated inspection results to Doug Hild, Site Vice President, and other members of the licensee staff.
- On March 26, 2024, the inspectors presented the inservice inspection results to Doug Hild, Site Vice President, and other members of the licensee staff.
- On March 29, 2024, the inspectors presented the radiation protection inspection results to Doug Hild, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Deaerator Tank Leak Rate Increase (Large Leak)
11/28/2023
OP-AA-108-111-
1001
Severe Weather and Natural Disaster Guidelines
QCOP 0010-01
Winterizing Checklist
Procedures
Seasonal Readiness
M-78
Diagram of Core Spray Piping
06/06/2005
Drawings
M-87
Diagram of High Pressure Coolant Injection - HPCI Piping
05/11/1998
QCOS 0020-02
Safety System Monthly Manual Valve Position Verification
QCOS 1400-10
Core Spray Operability Verification
QCOS 2300-10
HPCI Monthly Valve Position Verification
009
Procedures
QOM 2-1400-09
2A Core Spray Valve Checklist
HPCI Turbine/Pump, RES Oil Sample, Lube, and Inspection
05/17/2023
Sample 1B CS Motor Upper Bearing Oil
05/18/2023
HPCI Keep Fill Supply Check Valve Closure Test (IST)
01/24/2024
CCST/TORUS Level Switch Functional Test
11/30/2023
MM Replace 2A Core Spray Motor Cooling Water Hose
10/21/2023
OPS PMT 2-1402-A Core Spray Motor Cooling Lines
10/21/2023
Core Spray Pump A Flow Rate (IST)
2/28/2023
Work Orders
Safety System Manual Valve Position Verification
2/04/2023
NRC ID: U2 Cable Tunnel Disconnected Cable and Oil on
Floor
01/24/2024
Corrective Action
Documents
Quarterly Fire Brigade Drill Tracking / Trending Crew F
2/07/2024
Detection and Suppression Turbine Building Ground Floor
(Unit 2)
10/26/1998
Detection and Suppression Turbine Building Mezzanine
Floor (Unit 2)
05/02/2003
Drawings
M-27, Sheet 4
Diagram of Fire Protection Piping
04/17/1998
FZ 1.1.1.3
U1 Reactor Bldg El 623' Mezzanine Level
08/2022
FZ 1.1.2.3
U2 Reactor Bldg 623' Mezzanine Level
08/2022
FZ 120
Unit 1/2 SBO 595' Elev Station Blackout Building
08/2022
Fire Plans
FZ 6.2.B
U2 Turbine Bldg 615' ELev U2 DC Panel Room (Battery
08/2022
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Charger Room)
FZ 8.2.4
Unit 1 TB 580'-0" Elev. Cable Tunnel
08/2022
FZ 8.2.5
Unit 2 Cable Tunnel
08/2022
FZ 8.2.6.E
U2 Turbine Bldg 595" Elev Ground Floor Northern Area
08/2022
FZ 8.2.6.E
Unit 2 TB 595'-0" Elev. D Heater Bay
08/2022
FZ 8.2.7.E
Unit 2 TB 515'-6" Elev. North Mezzanine Floor
08/2022
FZ 8.2.7.E
Unit 2 TB 615'-6" Elev. North Mezzanine Floor
08/2022
FZ 8.2.8.C
U2 Turbine Bldg El 639' 4KV Bus 24-1 Switchgear Area
08/2022
Miscellaneous
Fire Drill Scenario
Information
Fire Drill Plan for the Unit 2 Trackway Expansion
ER-QC-600
Units 1 & 2 Site List of High Risk Fire Areas
OP-QC-201-012-
1001
Quad Cities On-Line Fire Risk Management
QCAP 1500-01
Administrative Requirements for Fire Protection
Procedures
RP-QC-837
Rad Protection SCBA Inspection Checklist
01/14/2024
(NEIL) Unit 2 Turbine Building Mezz Level Nozzle Inspection
06/27/2022
Turbine / Radwaste Building Fire Hose Reel Quarterly
Inspection
06/16/2022
(LR) Fire Damper Visual Inspection
06/29/2023
(NEIL) Turbine Building Mezz/Battery Room Fire Protection
Func Test
2/28/2023
(NEIL) Turb/Radwaste/LTD Bldg Fire Hose Reel Inspection
06/29/2023
Work Orders
Outside Building Hose Reel Fire Inspection
09/29/2023
Q2R26 IVVI Core Spray Pipe 4P4d Flaw
04/01/2022
Q2R26 IVVI FW Sparger Pins-Minor Wear/Crack
04/01/2022
Corrective Action
Documents
NOS QV ID Weld Reject Due to Linear Indications
04/04/2022
Engineering
Changes
Evaluation of Q2R26 IVVI Inspection Findings
Miscellaneous23-051
ASME Section XI Repair/Replacement Plan
08/22/2023
Surface Exam Report for 3001B-W-102A
03/25/2024
Surface Exam Report for 3001B-W-102A
03/25/2024
Volumetric Exam Report for 30C-S19
03/23/2024
NDE Reports
Volumetric Exam Report for N6A NOZ
03/28/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Volumetric Exam Report for N6A IRS
03/28/2024
Visual Exam Report for 3001B-W-102 A&B
03/22/2024
Magnetic Particle (MT) Examination
VT-3 Visual Examination of Component Supports,
Attachments and Interiors of Reactor Vessels
GEH-PDI-UT-1
PDI Generic Procedure for the Ultrasonic Examination of
Ferritic Welds
2.1
GEH-UT-300
Procedure for Manual Examination of Reactor Vessel
Assembly Welds in Accordance with PDI
Procedures
GEH-UT-311
Procedure for Manual Ultrasonic Examination of Nozzle
Inner Radius, Bore, and Selected Nozzle to Vessel Regions
Work Orders
Replace 2-1001-65B Low Pressure Discharge Elbow
09/22/2023
Corrective Action
Documents
CAPE 4722813
U1 EDG Failed During Performance of QCOS 6600-41
2/01/2024
Equipment Classification
GEK-39196
EJ Current-Limiting Power Fuses, EK Fuse Supports, and
Fuse-Disconnecting Switches
Procedures
QCEPM 0400-14
Emergency Diesel Generator Electrical Preventative
Maintenance
25
QDC-0200-N-
2453
Q2R27 Decay Heat and Related Calculations
Calculations
QDC-1900-N-
2454
Alternative Decay Heat Removal (ADHR) System
Qualification for Q2R27
Corrective Action
Documents
U1 EDG Output Breaker Tripped
01/11/2024
Engineering
Changes
Q2R27 Decay Heat and Related Calculations
Engineering
Evaluations
Evaluation in Support of RPV WIC Implementation
FZ 1.1.1.3
U1 Reactor Bldg, Elevation 623' Mezzanine Level
08/2022
FZ 1.1.2.2
Unit 2 RB 595' Elev. Ground Floor
08/2022
FZ 1.1.2.3
U2 Reactor Bldg. El 623' Mezzanine Level
08/2022
FZ 8.2.6.A
U1 Turbine Bldg El 595' 4KV Switchgear & U-1 Trackway
08/2022
Fire Plans
FZ 8.2.6.C
U1/2 Tubine Bldg, El 595' Ground Floor - Center Area
08/2022
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
FZ 8.2.7.A
Unit 1 Turbine Bldg, El 615'-6," Mezzanine Floor - Southern
Area
08/2022
FZ 8.2.7.B
Unit 1 TB 615'-6 Elev LP Heater Bay (East)/D Heater Bay
08/2022
FZ 8.2.7.C
U1/2 Turbine Bldg El 611'-6" Mezzanine Floor - Center Area
08/2022
FZ 8.2.7.D
U2 Turbine Bldg El 615' Mezzanine Floor (LP and D Heater
Bay)
08/2022
FZ 8.2.8.C
Unit 1/2 TB 639' Elev 480V Swgr 28/29 4KV Swgr 24-1
MG Set 2B
08/2022
Miscellaneous
Q2R27 Shutdown Safety Plan
03/05/2024
OP-QC-201-012-
1001
Quad Cities On-Line Fire Risk Management
Procedures
QCAN 901(2)-8
C-7
Diesel Generator 1(2) Fail to Start
Unexpected Alarm 901-5 D9, CHAH A Main Steam Tunnel
HI Temp
03/06/2024
2-1001-47 Calculated Closure Time Greater Than Design
11/08/2023
Extent of Condition Walkdown Actions on RHR, CS, and
11/10/2023
NRC Concern w/2BRHR Motor Cooling Line Not Having
Paperwork
2/21/2024
U2 SBO A Engine Governor Oscillating Erratically
2/28/2024
HPCI Switch Resistance Out of Tolerance per
AT# 00626169-03
2/27/2024
U1 HPCI Water Leaks from Turbine During Operations
2/28/2024
Follow-Up for HPCI Steam Leak Identified in IR 4754009
03/04/2024
AO 2-1301-12 Failed to Close Within Acceptable Limits
03/07/2024
U2 SBO A Engine Governor Oscillating Erratically
03/23/2024
QCAN 901(2)-5
D-9
Channel A Main Steam Line Tunnel High Temperature
Corrective Action
Documents
QCOS 1600-06
ECCS and Primary Containment Isolation Trip Instruments
Outage Report
Schematic Diagram PCI System Panel 901-15 Trip Logic
and Condenser
11/21/1998
Drawings
Schematic Diagram PCI System Panel 901-17 Trip Logic
06/14/2002
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
M-46
Diagram of High Pressure Coolant Injection Piping
W
M-89
Diagram of Reactor Core Isolation Cooling RCIC Piping
08/15/2003
Operability Evaluation of Rubber Hose to Motor Upper
Bearing Cooler of 2B RHR Motor
Engineering
Changes
ECR 460978
Engineering to Validate if the 304°F (151.1C) Temperature
Stated as the Worst-Case HELB Scenario is Applicable
When the 2-1001-47 Valve is Closed
11/28/2023
IM Perform Calibration on 1-2341-10
2/27/2024
Work Orders
U2 SBO A Engine Governor Oscillating Erratically
2/29/2024
Calculations
QDC-1400-M-
0559
Determination of Minimum Core Spray Pump Discharge
Pressure Required for Quarterly Pump Flow Rate Test
03/22/1998
FLEX Readiness Fleet Review
04/07/2017
EOID: RHR B Loop Min Flow Valve Not Declutched
08/13/2020
New Spring Pack for 1-1001-18B Needed
04/14/2022
EO ID: A FLEX Deep Well Pump Starter Panel Not Working
05/20/2023
Troubleshooting Results from Flex Deep Well Starter
2/09/2024
U1 HPCI Switch Resistance Out of Tolerance per
AT #00626169-032
2/27/2024
U1 HPCI Oil Leak from South Turbine Bearing During Run
2/28/2024
U1 HPCI Water Leaks from Turbine During Operation
2/28/2024
Flex Diesels Deficiencies
03/04/2024
NRC Question for Calibration Information
03/13/2024
Corrective Action
Documents
PSU: Q2R27 FW 2-0220-62A As-Found LLRT Exceeded
Action Limit
03/24/2024
Engineering
Changes
Guidance for IM Calibrate Valve Position for MOV 0-2901-6
Engineering
Evaluations
385320
Determination of Minimum Core Spray Pump Discharge
Pressure Required for Quarterly Pump Flow Rate Test
CC-QC-118-1003
Quad Cities Power Station Final Integrated Plan Document
Mitigating Strategies NRC Order EA-12-049
Primary Containment Leakrate Testing Program
Post Maintenance Testing
Procedures
Maintenance Alterations Process
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
QCOP 0050-07
FLEX Generator and Pump Power Cable Deployment
QCOS 0050-05
Flex Well Pump Test
QCOS 0100-07
Feedwater Check Valve Local Leak Rate Test
CK 1(2)-0220-58A/B, 1(2)-0220-62A/B
QCOS 0400-05
Mains Steam Isolation Valve Local Leak Rate Test
AO 1(2)-0203-1A/B/C/D, AO 1(2)-0203-2A/B/C/D
QCOS 1400-01
Core Spray System Flow Rate Test
QCOS 2300-05
HPCI Pump Operability Test
QCOS 2300-13
HPCI System Manual Initiation Test
QCOS 7500-05
SBGTS Operability Test
QCTP 0130-01
Primary Containment Leakrate Testing Program
QCOS 2900-01
Safe Shutdown Makeup Pump Flow Rate Test
HPCI GSC Hotwell Level Switch Replacement
2/26/2024
MM Repair Not Getting Full Flow out of 0-2901-17
01/10/2024
1/2-2901-6 MOV Local Indicator Discrepancy
07/24/2023
Calibrate Valve Position Indication on FIC 0-2940-7 and
FIC 0-2940-6
2/21/2024
HPCI Pump Discharge Flow Switch CAL/FUNC Test
2/26/2024
(IST)(NEIL) HPCI Pump Operability
2/29/2024
Work Orders
SBGT Operability (A Train)
01/24/2024
January 30th EP Pre-Exercise Results (Simulator)
2/01/2024
January 30th EP Pre-Exercise Results (Alternate Facility)
2/01/2024
Corrective Action
Documents
TQ-AA-113-F003
ERO Remedial Training Notification and Action on Failure
24-02-510
ERV/SRV/Target Rock Work Activities
03/11/2024
24-02-901
Reactor Disassembly/Reassembly Activities
03/11/2024
ALARA Plans
QC-02-24-00921-
Q2R27 - Rx Head Flange Repairs
Elevated Contamination Levels on RFF
03/18/2024
Corrective Action
Documents
Worker Contaminated Removing SRV Gaskets
03/24/2024
Miscellaneous
Beta and Alpha
Smear Results by
Area and System
Beta and Alpha Smear Results by Area and System - Unit 2
Outage Q2R27
03/26/2024
Radiation
2-24-213362
Refuel Floor
03/18/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
24-213342
Reactor Cavity
03/18/2024
RFF Back UP Air
Sample
Air Sample for RWP 905 2300 03182024
03/18/2024
RFF GA Back Up
Air Sample
Air Sample for RWP 905 2320 03182024
03/18/2024
RFF GA Back Up
Air Sample
Air Sample for RWP 905 2335 03182024
03/18/2024
RFF GA Back Up
Air Sample
Air Sample for RWP 905 2350 03182024
03/18/2024
U2 RFF NW
Cavity
Air Sample for RWP QC-2-24-00901 2300 03182024
03/18/2024
Surveys
Cavity
Air Sample for RWP QC-2-24-00901 2300 03182024
03/18/2024
QC-02-24-00510
Q2R27 DW Main Steam Safety Relief Valve Activities
QC-02-24-00510
Q2R27 DW Main Steam Safety Relief Valve Activities
QC-02-24-00901
Q2R27 FF RX Disassembly/Reassembly Activities
QC-02-24-00921
Q2R27 Refuel Floor Emergent Activities
Radiation Work
Permits (RWPs)
QC-02-24-00921
Q2R27 Refuel Floor Emergent Activities
Tritium Troubleshooting Results
09/25/2023
Corrective Action
Documents
5197731-01 Issues Putting U1 125 Batt. on Charger
2/06/2023
Radiological Groundwater Protection Program
Radiological Groundwater Protection Implementation
Program
Procedures
QCOS 6900-25
Transfer of Unit One 125 VDC Bus Between Normal and
Alternate Battery
RCIC Isolated During Performance of MA-QC-IM-1-13101
09/22/2023
Control Rod N-11 Scrammed In
2/09/2024
Corrective Action
Documents
OSP 2-1001-5A Bonnet Bolts Over Torqued
03/27/2024