IR 05000254/2025001
| ML25127A038 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 05/08/2025 |
| From: | Robert Ruiz NRC/RGN-III/DORS/RPB1 |
| To: | Rhoades D Constellation Energy Generation |
| References | |
| IR 2025001 | |
| Download: ML25127A038 (1) | |
Text
SUBJECT:
QUAD CITIES NUCLEAR POWER STATION - INTEGRATED INSPECTION REPORT 05000254/2025001 AND 05000265/2025001
Dear David Rhoades:
On March 31, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Quad Cities Nuclear Power Station and 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.
One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violation or the significance or severity of the 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 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 8, 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. 05000254 and 05000265 License Nos. DPR-29 and DPR-30 Enclosure:
As stated cc w/ encl: Distribution via LISTSERV Signed by Ruiz, Robert on 05/08/25
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
Failure to Establish HPCI Mission Time in Accordance with the Design Bases Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000254,05000265/2025001-01 Open/Closed
[H.14] -
Conservative Bias 71152A The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to assure the applicable regulatory requirements and the design basis for the safety-related high pressure coolant injection (HPCI) system were correctly translated into specifications, drawings, procedures, and instructions. Specifically, by establishing a 10-minute mission time for HPCI, the licensee failed to recognize the design bases as presented in the Updated Final Safety Analysis Report (UFSAR) included the requirement for HPCI to be capable of performing its specified safety functions without the support of the automatic depressurization system (ADS) system across a wide range of reactor pressures.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000265,07200053/2024-002-01 LER 2024-002-01 for Quad Cities Nuclear Power Station,
Unit 2, Turbine Trip and Automatic Scram due to Digital EHC Power Supply Intermittent Failure 71153 Closed
PLANT STATUS
Unit 1 The unit began the inspection period in its end-of-cycle coast down period. On March 10, 2025, the unit was shut down to commence refueling outage (RFO) Q1R28 where it remained until March 31, 2025, when the unit was restarted and began ascension to full power.
Unit 2 The unit began the inspection period at full-rated thermal power. On March 1, 2025, power was reduced to approximately 37 percent to perform corrective maintenance on the 2B adjustable speed drive. The unit returned to full-rated thermal power on March 2, 2025, where it remained at full-rated thermal power, with the exception of short-term power reductions for control rod sequence exchanges, 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 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
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (3 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
(1)1/2 emergency diesel generator (EDG) during a planned Unit 2 EDG maintenance window on January 29, 2025
- (2) Unit 2 EDG during a planned Unit 1 EDG maintenance window on February 5, 2025 (3)1/2 EDG cooling water system after planned maintenance on the 1/2 EDG on March 20, 2025
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (8 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) 1.1.1.5, Unit 1 reactor building, elevation 666'-6", standby gas treatment 4th floor east and standby liquid control 4th floor west on January 3, 2025
- (2) FZ 8.1, Unit 1/2 turbine building, elevation 595'-0", clean and dirty oil room on January 22, 2025
- (3) FZ 11.3.1, Unit 2 reactor building, elevation 554'-0", southwest corner room 2B core spray on January 15, 2025
- (4) FZ 3.0, services building, elevation 609'-0", cable spreading room on January 31, 2025
- (5) FZ 11.1.3, Unit 1 reactor building, elevation 554'-0", high pressure coolant injection (HPCI) and HPCI access tunnel on February 12, 2025 (6)hot work walkdown in the low-pressure heater bay during Q1R28 refueling outage on March 16, 2025
- (7) FZ 8.2.7.B, Unit 1, elevation 615'-6", low pressure heater bay (east) / D heater bay on March 18, 2025
- (8) FZ 11.1.1 B, Unit 1 turbine building, elevation 547'-0", residual heat removal (RHR)service water pump on March 20, 2025
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the onsite fire brigade training and performance during an unannounced fire drill on February 12, 2025.
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (2 Samples)
- (1) The inspectors observed and evaluated licensed operator performance in the control room during a normal unit shutdown for RFO Q1R28 on March 10, 2025.
- (2) The inspectors observed and evaluated licensed operator performance in the control room during a normal unit startup from RFO Q1R28 on March 31, 2025.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated an as-left crew performance evaluation on February 6, 2025.
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) 4825109, EO ID 1-1402-25A Local Stop Pushbutton Does Not Release, on January 21, 2025
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (5 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)work management risk for the work week of January 27, 2025, on January 27, 2025 (2)shutdown safety plan review for RFO Q1R28 on February 27, 2025 (3)shutdown safety and online fire risk equipment walkdowns during RFO Q1R28 for the week of March 10, 2025 (4)shutdown safety and online fire risk equipment walkdowns during RFO Q1R28 for the week of March 17, 2025 (5)shutdown safety and online fire risk equipment walkdowns during RFO Q1R28 for the week of March 24, 2025
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (9 Samples)
The inspectors evaluated the licensees justifications and actions associated with the following operability determinations and functionality assessments:
- (1) AR 4825476, Leak Identified on 1/2 EDGCWP [emergency diesel generator cooling water pump] Piping Upstream of 0-3999-139, on January 6, 2025
- (2) AR 4827645, AO-2-1601-58 Closed Immediately Upon Opening, on January 8, 2025
- (3) AR 4824047, Did not Receive Expected Alarm at Alarm Setpoint, on January 14, 2025
- (4) AR 4831512, NRC/IEMA [Illinois Emergency Management Agency] Questions Regarding EDG Air Line Lubricator, on January 22, 2025
- (5) AR 4833062, U1 SBO [station blackout] 1B Jacket Water Booster Pump Failed to Auto-Start, on January 30, 2025
- (6) AR 4838330, QCOS 2300-05 Acceptance Criteria Not Met Due to Miscalibrated FT, on February 18, 2025
- (7) AR 4844777, Q1R28:MOV 1-1001-19A Underthrust, on March 18, 2025
- (8) AR 4848496, OSP-1/2 EDG Vent Fan Started from ALT Feed - QCOS 6600-43, on March 27, 2025
- (9) AR 4847782, PSU - Received 902-3 F-14 HPCI Lo Flow and MGU Not at HSS, on March 24, 2024
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)
The inspectors evaluated the following temporary or permanent modifications:
- (1) Work Order (WO) 5373949, Temporary Installation of Reactor Vessel Level Upper 400 Reference Leg and Local Level Gauge during Q1R28 Refueling Outage, on March 10, 2025
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Partial)
(1)
(Partial) The inspectors evaluated Q1R28 RFO activities from March 9 through March 31, 2025.
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) (10 Samples)
- (1) WO 5609605, Leak Identified on 1/2 EDGCWP Piping Upstream of 0-3999-139, on January 6, 2025
- (2) WO 5587947, Pressure Suppression Valve Timing Test (IST), on January 8, 2025
- (3) AR 4827694, Closure of Operations PMTs, on January 30, 2025
- (4) Unit 1 EDG following planned maintenance work window on February 6, 2025 (5)as-left Unit 1 2A main steam isolation valve local leak rate testing (LLRT) on March 25, 2025
- (6) LLRT following WO 4802725, Feedwater Check Valve Disassemble/Inspection, on March 25, 2025
- (7) PMT following repairs associated with AR 4845470 PSU 1-0220-59A Unable to Close during QCOS 3200-04, on March 25, 2025
- (8) LLRT following WO 5343759, Rebuild Actuator and Replace Solenoid, on March 31, 2025
- (9) WO 5377897, Reactor Vessel Class 1 and Associated Class 2 System Leak Test, on March 26, 2025
- (10) WO 5354834, Scram Timing During Reactor Hydrostatic Testing, on March 26, 2025
Surveillance Testing (IP Section 03.01) (2 Samples)
- (1) QCOS 6600-51, Unit 1 Emergency Diesel Generator Start Failure Logic Test, on February 3, 2025
- (2) QCOS 6600-37, Unit 1 Div 1 Emergency Diesel Generator Largest Load Reject Surveillance, on March 26, 2025
Inservice Testing (IST) (IP Section 03.01) (3 Samples)
(1)2B standby liquid control flow rate test on January 2, 2024
- (2) QCOS 1600-50, 18-inch Primary Containment Vent and Purge Air Operated Valve Surveillance, on February 25, 2025
- (3) QCOS 0250-04, MSIV [main steam isolation valve] Closure Timing, on March 11, 2025
Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)
- (1) QCOS 0100-05, Main Steam Isolation Valve Local Leak Rate Test AO 1(2)-0203-1A/B/C/D, AO 1(2)-0203-2A/B/C/D, on March 10, 2025
71114.06 - Drill Evaluation
Required Emergency Preparedness Drill (1 Sample)
(1)off-year exercise on February 4, 2024
Additional Drill and/or Training Evolution
The inspectors evaluated:
(1)emergency preparedness drill on January 14,
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls
Radiological Hazard Assessment (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards.
Instructions to Workers (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated 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 the main control point during a refueling outage.
- (2) Licensee surveys of potentially contaminated material leaving the RCA and workers exiting the RCA at the trackway 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) Unit 1 inboard MSIV overhaul
- (2) Unit 1 drywell nuclear instrumentation system
- (3) Unit 1 torus diving 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 (HRAs)and very high radiation areas (VHRAs):
- (1) Unit 1 drywell access control point
- (2) Unit 2 traverse incore probe room
- (3) Unit 1 clean up 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
Permanent Ventilation Systems (IP Section 03.01) (1 Sample)
The inspectors evaluated the configuration of the following permanently installed ventilation systems:
(1)standby gas treatment system (SBGTS)
Self-Contained Breathing Apparatus for Emergency Use (IP Section 03.04) (1 Sample)
- (1) The inspectors evaluated the licensees use and maintenance of self-contained breathing apparatuses.
71124.04 - Occupational Dose Assessment
Source Term Characterization (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated licensee performance as it pertains to radioactive source term characterization.
External Dosimetry (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee processes, stores, and uses external dosimetry.
Internal Dosimetry (IP Section 03.03) (2 Samples)
The inspectors evaluated the following internal dose assessments:
(1)whole body count and dose calculations for intake on March 26, 2024 (2)whole body count and dose calculations for intake on April 2, 2024
Special Dosimetric Situations (IP Section 03.04) (2 Samples)
The inspectors evaluated the following special dosimetric situations:
(1)use of extremity dosimetry (finger and foot rings) to measure shallow dose equivalent (SDE) for reactor head disassembly and reassembly (2)use of extremity dosimetry (finger and foot rings) to measure SDE for reactor cavity drain down
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, 2024, through December 31st, 2024)
- (2) Unit 2 (January 1, 2024, through December 31st, 2024)
IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (2 Samples)
- (1) Unit 1 (January 1, 2024, through December 31st, 2024)
- (2) Unit 2 (January 1, 2024, through December 31st, 2024)
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (2 Samples)
- (1) Unit 1 (January 1, 2024, through December 31, 2024)
- (2) Unit 2 (January 1, 2024, through December 31, 2024)
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)extent-of-condition review related to the use of non-conservative motor operated valve factors as documented in NRC Comprehensive Engineering Team Inspection Report 05000254/2023010 and 05000265/2023010 dated January 23, 2025 (2)design basis justification for a HPCI 10-minute mission time on March 26, 2025
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) Licensee Event Report (LER) 265/2024-002-00, Turbine Trip and Automatic Scram due to Digital EHC Power Supply Intermittent Failure (Agencywide Document Access and Management System (ADAMS) Accession No. ML24304A856). The inspectors determined that the cause of the condition described in the LER was not reasonably within the licensees ability to foresee and correct and therefore was not reasonably preventable. No performance deficiency nor violation of NRC requirements was identified. This LER is closed.
INSPECTION RESULTS
Failure to Establish HPCI Mission Time in Accordance with the Design Bases Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000254,05000265/2025001-01 Open/Closed
[H.14] -
Conservative Bias 71152A The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to assure the applicable regulatory requirements and the design basis for the safety-related high pressure coolant injection (HPCI) system were correctly translated into specifications, drawings, procedures, and instructions. Specifically, by establishing a 10-minute mission time for HPCI, the licensee failed to recognize the design bases as presented in the Updated Final Safety Analysis Report (UFSAR) included the requirement for HPCI to be capable of performing its specified safety functions without the support of the automatic depressurization system (ADS) system across a wide range of reactor pressures.
Description:
During the Age-Related Degradation Inspection (ARDI), the inspectors reviewed the HPCI gland seal condenser (GSC). The GSC and its subcomponents are safety-related and prevent steam used to operate the HPCI turbine from leaking into the HPCI room.
The inspectors noted the licensee requires the GSC to function for the HPCI System to be considered available to perform its probabilistic risk assessment function(s). However, the licensee did not credit the GSCs function when determining HPCI operability because the licensee believed the HPCI System only needed to operate for 10 minutes to perform its Technical Specification (TS) specified safety function (SSF). As a result, the licensee would consider the HPCI System to be operable but not available when the GSC was out of service.
The inspectors discussed the HPCI 10-minute mission time with the licensee as part of the ARDI. During these discussions, the licensee stated the basis for the 10-minute mission time was discussed in UFSAR Section 6.3.3.1.3.2, Evaluation of Subsystem Performance, which states:
The HPCI turbine oil cooler and gland seal condenser are cooled by water from the suppression pool. Since these components are rated at 140 °F, continued operation above a suppression pool temperature of 140 °F is not permitted. Also, operation of HPCI above 140 °F would exceed the current net positive suction head (NPSH) calculations for rated HPCI pump flows. Another limitation on the HPCI system is related to the dependence of the HPCI room cooler on the unit emergency diesel generator (EDG). Therefore, any single failures of the unit EDG need to assume consequential loss of the HPCI system after 10 minutes of operation. As a result of these considerations, the HPCI system is not credited when any of these conditions are exceeded. The results of the analysis show that the HPCI system met its requirements before the 10-minute mission time was exceeded and the suppression pool temperature exceeded 140 °F (see Reference 84 for AREVA analysis).
The inspectors reviewed historical documents and determined the HPCI 10-minute mission time was added to the UFSAR in approximately 2006 via Engineering Change (EC) 349583, Revision 1, Change Needed to Support Initial Use of Westinghouse Optima 2 Fuel in Unit 1.
The inspectors noted the 10-minute mission time conflicted with the design bases of the HPCI System. Of particular interest, the design bases discussed in UFSAR Chapter 6, Chapter 15, and the TS Bases describe that the emergency core cooling systems (ECCS)are evaluated for the entire spectrum of break sizes. The HPCI System is responsible for small breaks by providing core cooling for a wide range of reactor pressures (150 psig to 1120 psig). The available descriptions of the HPCI design bases indicate HPCI is expected to provide core cooling by itself or with the support of one of the low pressure ECCS systems (i.e., core spray and/or low-pressure coolant injection). Under no design bases event, was HPCI described as working in combination with the ADS. This is because ADS serves as a backup to HPCI and is not intended to work in combination with HPCI in order to perform a design bases function.
During conversations with licensee staff, the inspectors found that the licensee was crediting using a combination of HPCI and ADS to address some small break loss of coolant accidents (LOCAs) or loss of feedwater transients. Additionally, it was noted for certain small breaks, if using HPCI, the reactor coolant system (RCS) would take longer than 10 minutes to reach the injection setpoint of the low pressure ECCS. Based on the information provided above, the inspectors concluded the HPCI 10-minute mission time appears to be mainly driven by the suppression pool water reaching and exceeding the evaluated temperature of 140 F and impacting the HPCI net positive suction head (NPSH)analysis, the oil cooler rating, and the GSC rating. Since classifying a system as operable but not available and using HPCI and ADS together to mitigate certain design basis events is not typically seen within the nuclear industry, the inspectors asked the licensee to clarify the basis for using a 10-minute mission time for the HPCI System.
On October 9, 2024, the licensee provided Engineering Change (EC) 642418, Documentation of the Basis for the HPCI 10 Minute Mission Time, Revision 0 to the NRC.
Within this document, the licensee stated the HPCI 10-minute mission time was established via EC 357611 dated October 28, 2005. EC 642418 also discussed the licensees position regarding the appropriateness of the 10-minute mission time based upon the following factors:
1. Defining SSF and mission time
2. Determining the HPCI System SSF(s) per applicable documents
3. Using both HPCI and the ADS to mitigate pipe breaks discussed in Chapters 6 and 15
of the UFSAR and the TS Bases
4. The 10-minute mission time aligns with the 10 CFR 50.36(c)(2)(ii) criterion as
discussed in the Agencys July 16, 1993, document, NRCs Final Policy Statement on Technical Specification Improvement Defining SSF and Mission Time The inspectors noted the Quad Cities TS define the term operable/operability as:
A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
As expected, the key term in this definition is SSF. Although the TS do not define SSF, this term is defined in both NEI 18-03, which is not endorsed by the NRC but was used by the licensee in EC 642418, and IMC 0326.
NEI 18-03 defines SSF as:
A SSF is a function of controlling importance to safety assumed to be performed by a system, structure, or component (SSC) in the analyses and evaluations summarized in the UFSAR, typically Chapters 6 and 15, or the plant-specific equivalent chapters. These SSFs are needed to obviate the risk of an immediate threat to the public health and safety. SSFs are the subset of all SSC functions that meet one or more criterion in 10 CFR 50.36(c)(2)(ii), as described in the NRCs Final Policy Statement on TS Improvement, unless otherwise stated in the docketed plant-specific SSF scope. Consequently, the SSFs may not be all of the SSC functions described in the UFSAR. The plant-specific SSF scope derives from the functions and design conditions for performance relied on by the licensee and the NRC when the TS were prepared, submitted, reviewed, and approved. For plants with Standard Technical Specifications, these functions are typically discussed in the TS Limiting Condition for Operation (LCO) Bases.
NEI 18-03 also defines mission time as:
the time an SSC must be capable of performing the SSF Comparatively, NRC IMC 0326 defines SSF as follows:
The specified function/SSF of an SSC is that specified safety function(s) in the CLB [current licensing basis] for the facility. Not all SSC functions described in the CLB are SSFs required for operability as described in Section 03.01.b.
Section 03.01.b states:
SSCs that are not explicitly required to be operable by TS but perform necessary and related support functions for TS SSCs are required to be operable by TS. SSCs may also have design functions that do not perform a necessary and related support function for TS SSCs. These design functions are not within the scope of an OD [operability determination].
The inspectors concluded that NEI 18-03 and NRC IMC 0326 agree that not all SSC functions described in the UFSAR are SSFs. However, the inspectors found the licensee has interpreted this information to conclude certain SSCs design bases functions are not SSFs.
The inspectors disagreed with the licensees position since it was inconsistent with the definition of design bases provided in 10 CFR 50.2, the IMC 0326 guidance, and the NEI 18-03 definition of SSF.
Determining the HPCI System SSF(s) Using Applicable Documents
Section 50.2 of 10 CFR defines the term design basis as follows:
that information which identifies the specific functions to be performed by a structure, system or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values may be:
- (1) restraints derived from generally accepted state of the art practices for achieving functional goals, or
- (2) requirements derived from analysis (based on calculation and/or experiments)of the effects of a postulated accident for which a structure, system or component must meet its functional goals.
Both NEI 18-03 and IMC 0326 use the 50.2 definition of design bases provided above and add the following statement to the definition:
The design basis for safety-related SSCs is initially established during the original plant licensing and related primarily to accident and event prevention or mitigation functions.
The inspectors found additional design bases guidance in NEI 97-04, Appendix B, Guidance and Examples for Identifying 10 CFR 50.2 Design Basis, which was endorsed by the NRC via Regulatory Guide 1.186 and referenced by both NEI 18-03 and IMC 0326.
Specifically, NEI 97-04, Appendix B, states:
10 CFR 50.2 design bases consist of the following:
Design bases functions: Functions performed by SSCs that are
- (1) required by, or otherwise necessary to comply with, regulations, license conditions, orders or technical specifications, or
- (2) credited in licensee safety analyses to meet NRC requirements.
As a result, the inspectors concluded the definition and available guidance on design bases and design bases function was centered on specific functions to be performed by an SSC to stay within the bounds of the design, requirements, TS, etc. This language is similar to the wording in NEI 18-03 which states that the plant-specific SSC scope is derived from the functions and design conditions for performance relied upon by the licensee and the NRC when the TS were prepared, submitted, reviewed and approved.
The inspectors continued to compare the HPCI design bases information with the SSF definition provided in NEI 18-03. Specifically, the inspectors focused on the statements within NEI 18-03 which indicated the SSFs are typically discussed in the TS Bases for plants with Standard Technical Specifications, such as Quad Cities, and Chapters 6 and 15 of the UFSAR. Quad Cities TS Bases Section B 3.5.1, ECCS-Operating, states the following regarding the SSFs of the HPCI and ADS systems:
From the background portion of Section B 3.5.1:
[Referring to ADS] Although the system is initiated, ADS action is delayed, allowing the operator to interrupt the timed sequence if the system is not needed.
The HPCI pump discharge pressure almost immediately exceeds that of the RCS [reactor coolant system], and the pump injects coolant into the vessel to cool the core. If the break is small, the HPCI System will maintain coolant inventory as well as vessel level while the RCS is still pressurized. If HPCI fails, it is backed up by ADS in combination with LPCI [low pressure coolant injection]
and CS [core spray]. In this event, the ADS timed sequence would be allowed to time out and open the relief valves and safety/relief valve (S/RV) depressurizing the RCS, thus allowing the LPCI and CS to overcome RCS pressure and inject coolant into the vessel.
The HPCI System is designed to provide core cooling for a wide range of reactor pressures (150 psig to 1120 psig).
The ADS consists of 5 valves (4 relief valves and one safety relief valve). It is designed to provide depressurization of the RCS during a small break LOCA if HPCI fails or is unable to maintain required water level in the RPV.
The Applicable Safety Analysis section of Section B 3.5.1 states:
The ECCS performance is evaluated for the entire spectrum of break sizes for a postulated LOCA. The accidents for which ECCS operation is required are presented in References 5 [UFSAR Section 15.6.4.] and 6 [UFSAR Section 15.6.5.]. The required analyses and assumptions are defined in Reference 7
[10 CFR 50, Appendix K]. The results of these analyses are also described in Reference 8 [UFSAR Section 6.3.3.].
The inspectors reviewed UFSAR Section 15.6.4, Steam System Line Break Outside Containment, and Section 15.6.5, Loss-of-Coolant Accidents Resulting from Piping Breaks Inside Containment. The inspectors found UFSAR Subsection 15.6.4.4, Barrier Performance, discusses HPCI initiation approximately 40 seconds after an accident to provide additional continuous core cooling. Subsection 15.6.5.2, Sequence of Events and System Operation, states the HPCI, residual heat removal (RHR), and core spray systems would act to cool the core following LOCAs inside containment. Additionally, this same subsection stated the following:
For breaks in small liquid lines up to about 0.12 ft2 in area, HPCI can supply sufficient coolant to depressurize the vessel and cool the core, depending only on the core spray system and the LPCI mode of RHR for long-term cooling.
For breaks in liquid lines between 0.12 ft2 and 0.2 ft2 in area, the depressurizing function of the HPCI and the coolant makeup function of either the core spray subsystem or the LPCI mode of RHR would act in conjunction to provide effective core cooling. In the event of a LOCA without HPCI capability (i.e., if the normal feedwater and HPCI are assumed to be unavailable), the ADS would cause the reactor vessel blowdown to occur in a time interval sufficiently short to permit operation of the core spray system and LPCI mode of RHR to assure adequate core cooling.
The inspectors also reviewed UFSAR Section 6.3.3.1, ECCS Performance Evaluation, and found several examples reaffirming the design bases and the SSFs of the HPCI system are to work in conjunction with the LPCI or CS systems to provide the depressurization and core cooling functions for the entire spectrum of small breaks while the ADS serves as an independent backup to the HPCI system.
Lastly, the inspectors reviewed General Electric (GE) Report GE-NE-0000-0036-4362-01, Small Break LOCA - Dresden and Quad Cities Maximum Suppression Pool Temperature during HPCI Operation, dated September 30, 2005, and found that GE had recognized and quantified some small break LOCAs which would require the HPCI system to operate for longer than 10 minutes to allow the RCS to depressurize to the point where the LPCI or CS systems could inject water into the reactor core. Based upon the above UFSAR, TS Bases, and GE information, the inspectors concluded the SSF of the HPCI system includes the ability to depressurize the RCS and provide cooling of the reactor core for the full spectrum of small pipe breaks including those breaks which would require HPCI Operation for greater than 10 minutes.
Use of HPCI and ADS to Mitigate Small Pipe Breaks Within EC 642418, the licensee stated the following:
The design basis function of HPCI is discussed later in the UFSAR in section 6.3.1.3. This section states the HPCI subsystem when combined with the remaining ECCS after a single failure is provided to ensure that adequate core cooling takes place for all break sizes as directed by 10 CFR 50, Appendix K single failure ECCS analysis requirements.
The inspectors agree UFSAR Section 6.3.1.3 states the HPCI system can be combined with other ECCS to ensure adequate core cooling takes place for all break sizes. However, the inspectors found UFSAR Sections 6.3.3.1 and 15.6.5.2 are specific regarding which ECCS (low pressure coolant injection and core spray) work with HPCI to ensure core cooling.
The inspectors also found the use of ADS is discussed only when the HPCI system fails.
As part of EC 642418, the licensee also stated:
While ADS is designed to be an alternative to HPCI, this is to be considered a design basis function and not a Specified Safety Function. HPCI and ADS may be used together, while adhering to the single failure criteria, to support the ECCS Specified Safety Function of maintaining the fuel peak cladding temperature (PCT) within the acceptable limit during a LOCA as demonstrated by the ECCS LOCA analysis.
Based on conversations with licensee staff and the quoted EC 642418 information above, the inspectors concluded the licensee position was that the SSF of HPCI is to mitigate those break(s) which result in the most limiting PCT. However, this position was contrary to the definition of SSF provided in NEI 18-03 and NRC IMC 0326 which tie the definition of SSF, and TS operability, to Chapters 6 and 15 of the UFSAR and the TS Bases. As discussed previously, the UFSAR and TS Bases indicate HPCI must operate for the full spectrum of small pipe breaks regardless of the corresponding PCT.
Review of 10 CFR 50.36(c)(2)(ii) Criterion and NRC Policy Statement Document The NEI 18-03 definition of SSF discusses that SSFs are the subset of all SSC functions that meet one or more of the criteria in 10 CFR 50.36(c)(2)(ii), as described in the NRCs July 16, 1993, document, NRCs Final Policy Statement on Technical Specifications Improvement (ML20056F543). The Quad Cities TS Bases also acknowledges the HPCI system falls under Criterion 3 of 10 CFR 50.36(c)(2)(ii) which states:
Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
The reference policy statement provides the following additional discussion on Criterion 3:
A safety sequence analysis is a systematic examination of the actions required to mitigate the consequences of events considered in the plant design basis accident and transient analysis, as presented in chapter 6 and 15 of the plants FSAR. Such a safety sequence analysis considers all applicable events whether explicitly or implicitly presented. The primary success path of a safety sequence analysis consists of the combination and sequence of equipment needed to operate (including consideration of the single failure criteria), so that the plant response to Design Basis Accidents and Transients limit the consequences of these events to within the appropriate acceptance criteria.
As previously discussed, the design bases described in Chapters 6 and 15 of the Quad Cities UFSAR present a series of actions (i.e., the SSF(s)) the HPCI system must be capable of performing to mitigate the consequences of events included in the plants design bases, accidents, and transient analyses. These actions are consistently described within the UFSAR as being independently performed by HPCI and without dependence upon the ADS.
As a result, the inspectors concluded the SSF of the HPCI system includes the ability to mitigate the full spectrum of small line breaks discussed in UFSAR Chapters 6 and 15 and the TS Bases. The ADS system is only used in cases where HPCI fails. Additionally, the inspectors determined the licensees establishment of a HPCI 10-minute mission time was incorrect since this mission time did not include the mitigation of the full spectrum of design basis small pipe breaks which required HPCI Operation.
Corrective Actions: The licensee entered the inspectors concern into the corrective action program in order to evaluate it.
Corrective Action References: AR 4797078, NRC ARDI ID: HPCI 10 Minute Mission Time Basis.
Performance Assessment:
Performance Deficiency: The licensee failed to recognize the 10-minute mission time evaluated and accepted in EC 357611 would not bound all the small break LOCAs and losses of feedwater transients HPCI is required to be capable to mitigate without the support from the automatic depressurization system (ADS). As a result, the 10-minute HPCI mission time was contrary to the design bases and SSFs of the HPCI system as discussed in UFSAR Chapter 6, Chapter 15 and the TS Bases. Additionally, this was contrary to 10 CFR Part 50 Appendix B, Criterion III, Design Control, and is 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, accepting a 10-minute mission time for HPCI could allow the licensee to incorrectly conclude the HPCI system was operable during configurations and/or degraded conditions which would physically prevent HPCI from running more than 10 minutes. As a result, the inspectors concluded a detailed risk evaluation (DRE) was needed since this performance deficiency constituted a design issue that could result in the loss of operability/functionality of the HPCI system.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power.
A senior reactor analyst (SRA) performed a DRE using SAPHIRE, Version 8.2.11, and the Quad Cities Standardized Plant Analysis Risk model, Version 8.82, to determine the significance of the finding. The SRA used an exposure time of 1 year, which is the maximum permissible exposure time per the Risk Assessment of Operation Events (RASP) Handbook, Volume 1, Section 2.7 (ML17348A149).
The SRA modeled the finding as a change in the HPCI mission time. The SRA calculated the difference in the core damage frequency assuming a mission time of 10 minutes and the core damage frequency assuming a mission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for transients and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for all other accident scenarios, which corresponds to the maximum HPCI mission time provided in QC-PSA-005.06, Revision 7, High Pressure Coolant Injection System (HPCI) Notebook.
The senior reactor analyst did not modify any other mitigation capability or recovery credit.
The dominant core damage sequences for this finding involved
- (1) a small break LOCA, failure of the feedwater system, failure of the HPCI system, and failure to depressurize the reactor with the ADS, and
- (2) a weather-related loss of offsite power, failure of the emergency diesel generators, failure to recover offsite power and the emergency diesel generators, failure of the reactor core isolation cooling system, and failure of the HPCI system.
The SRA determined the change in core damage frequency to be less than 1E-7 per year.
Therefore, the SRA did not consider the risk contribution from external events or the change in large early release frequency. The SRA determined the finding was of 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. Although the performance deficiency occurred more than 3 years ago, the decision-making practices which resulted in establishing the HPCI 10-minute mission time were considered reflective of current performance as provided in Section 3.14a of Inspection Manual Chapter 0612, Issue Screening. Specifically, the NRC questioned the acceptability of the HPCI 10-minute mission time less than 2 years ago during a review of HPCI gland seal condenser pressure control valve issues. However, the licensee took minimal action to re-review the mission time issue to ensure its continued acceptability.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in § 50.2 and as specified in the license application, for those systems to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions.
UFSAR Section 6.3.2.3, High Pressure Coolant Injection Subsystem, summarizes the established design bases events and accidents used to design the HPCI system and states, The HPCI subsystem is designed to pump water into the reactor vessel under LOCA conditions which do not result in rapid depressurization of the pressure vessel. The loss of coolant might be due to a loss of reactor feedwater or to a small line break which does not cause immediate depressurization of the reactor vessel.
UFSAR Section 15.6.4.4, Barrier Performance, which is part of UFSAR Chapter 15.6.4 Steam System Line Break Outside Containment, states, The maximum MSIV closure time of 10.5 seconds would limit the total amount of liquid and steam lost from the primary system to prevent the core from being uncovered. Therefore, continuous cooling of the reactor core would be maintained throughout the transient and the subsequent initiation of the HPCI system would provide additional continuous core cooling.
UFSAR Section 15.6.5.2, Sequence of Events and Systems Operation, which is part of UFSAR Chapter 15.6.5 Loss-of-Coolant Accidents Resulting from Piping Breaks Inside Containment, states, The HPCI, RHR, and core spray systems would act to cool the core following the accident. For breaks in small liquid lines up to about 0.12 ft2 in area, HPCI can supply sufficient coolant to depressurize the vessel and cool the core, depending only on the core spray system and the LPCI mode of RHR for long-term cooling. For breaks in liquid lines between 0.12 ft2 and 0.2 ft2 in area, the depressurizing function of the HPCI and the coolant makeup function of either the core spray subsystem or the LPCI mode of RHR would act in conjunction to provide effective core cooling. In the event of a LOCA without HPCI capability (i.e., if the normal feedwater and HPCI are assumed to be unavailable), the ADS would cause the reactor vessel blowdown to occur in a time interval sufficiently short to permit operation of the core spray system and LPCI mode of RHR to assure adequate core cooling.
The licensee established EC 357611, HPCI Inputs Validated for LOCA Analysis, Revision 0, as the engineering product used to ensure the site remained within its design bases. This EC documented the basis for the licensee concluding that the HPCI system could perform its specified safety function(s) by operating for 10 minutes.
Contrary to the above, from October 28, 2005 until March 26, 2025, the licensee failed to assure the applicable regulatory requirements and the design basis for the safety-related HPCI system were correctly translated into specifications, drawings, procedures, and instructions. Specifically, by establishing a 10-minute mission time for HPCI, EC 357611 failed to recognize the design bases as presented in UFSAR Sections 6.3.2, 15.6.4 and 15.6.5 included the requirement for HPCI to be capable of performing its specified safety functions without the support of the ADS system across a wide range of reactor pressures (150 psig to 1120 psig). These SSFs included the ability to mitigate a spectrum of LOCA conditions which do not result in rapid depressurization of the pressure vessel. These losses of coolant might be due to a loss of reactor feedwater or to a small line break which does not cause immediate depressurization of the reactor vessel.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Observation: Extent-of-condition Review Related to the Sites Use of Non-conservative Motor Operated Valve Factors 71152A The inspectors performed a detailed review of Issue Report (AR) 04714457, NRC CETI:
[Motor Operated Valve] MOV Valve Factor Below Basis Value, and AR 04733415, NRC NCV 2023-010 Non-Conservative Valve Factors. Specifically, during the 2023 NRC Comprehensive Engineering Team Inspection, the inspectors identified the licensee was using a non-conservative valve factor from industry testing to size and set-up the 1-1001-23A MOV, instead of using the bounding valve-specific, empirically determined valve factor.
The inspectors focused on the licensees corrective actions and extent-of-condition review of the remaining susceptible valves under the Generic Letter (GL) 96-05 MOV Program.
The licensee reviewed 116 gate valve design datasheets to verify the bounding valve-specific valve factors were utilized to size and set-up the MOVs. The licensee revised 69 gate valve datasheets to incorporate the applicable valve factors from the applicable design basis document, Nuclear Engineering Standard (NES) NES-MS-06.6, MOV Valve Factors, Revision 2. As noted in AR 04716301, MOV 1-1001-26A Valve Factor Revision and Unacceptable Margin the licensee performed an operability evaluation (EC 640236)to restore positive margin on the 1-1001-26A valve.
The inspectors sampled MOV datasheets to verify the licensee had updated the design calculations to utilize the applicable bounding valve factors. Additionally, the inspectors reviewed licensee procedure, ER-AA-302-1001, MOV Rising Stem Motor Operated Valve Thrust and Torque Sizing and Set-up Window Determination Methodology, Revision 14, to review the methodology used to assure proper sizing and set-up windows for MOVs.
The inspectors did not identify any issues with the corrective actions taken.
No findings or violations of NRC requirements were identified by the inspectors in the course of this review.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified that no proprietary information was retained or documented in this report.
- On March 31, 2025, the inspectors presented the integrated inspection results to Doug Hild, Site Vice President, and other members of the licensee staff.
- On March 20, 2025, the inspectors presented the radiation protection inspection results to Kent Akre, Radiation Protection Manager, and other members of the licensee staff.
- On March 26, 2025, the inspectors presented the HPCI mission time review debrief inspection results to Conner Bealer, Regulatory Assurance Manager, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
QCOS 0020-02
SAFETY SYSTEM MONTHLY MANUAL VALVE POSITION
VERIFICATION
QCOS 6600-43
Unit 1/2 Emergency Diesel Generator Load Test
QOM 0-6600-01
Unit 0 Diesel Generator Valve Checklist
Procedures
QOM 2-6600-01
UNIT 2 DIESEL GENERATOR VALVE CHECKLIST
Calculations
QDC-4100-M-0691
Combustible Loading Calculation for the Power Block, SBO
Building and Crib House
HPCI Interlock Penetration Degraded
01/28/2025
Corrective Action
Documents
Exp Scope for WO# 5622996-01
(HPCI Wall Penetration U-1)
2/11/2025
Corrective Action
Documents
Resulting from
Inspection
Scaffolds Not Properly Tracked
03/21/2025
Drawings
M-27, Sheet 2
Diagram of Fire Protection Piping
09/10/1999
Engineering
Changes
Implement Fire Hose Removal Strategy
FZ 1.1.1.5
Unit 1 RB 666'-6" Elev. Stand-By Gas Treatment 4th Floor
East
Unit 1 RB 666'-6" Elev. Stand-By Liquid Control 4th Floor
West
08/2022
FZ 11.1.1 B
Unit 1 TB 547'-0" Elevation RHR Service Water Pumps
08/2022
FZ 11.3.1
Unit 2 RB 544'-0" Elev. SW Corner Room - 2B Core Spray
08/2022
FZ 3.0
SB 609'-0" Elev. Cable Spreading Room
08/2022
Fire Plans
FZ 8.1
Unit 1/2 TB 595'-0" Elev. Clean and Dirty Oil Room
08/2022
FZ 8.2.7.B
Unit 1 TB 615'-6" Elev. LP Heater Bay (East)/D Heater Bay
08/2022
SCAFFOLD INSTALLATION, MODIFICATION, AND
REMOVAL REQUEST PROCESS
Fire Drill Performance
Procedures
QCAP 1500-01
Administrative Requirements for Fire Protection
QCGP 1-1
Normal Unit 1 Startup
2
Procedures
QCGP 2-1
Normal Unit Shutdown
109
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
EO ID 1-1402-25A Local Stop Pushbutton Does
Not Release
2/20/2024
Engineering
Changes
Design Consideration Change Summary for Local Motor
Operated Valve Pushbutton Stations
SBO DG1 Engine B L/O Temp Low Alarm Received
01/27/2025
2D RHRSW Vault Wall Plug Will Not Reattach
01/28/2025
QCOS 1000-04 Vibe Collection Inadvertently N/Ad
01/30/2025
U2 EDG Oil Fill Issue
01/30/2025
Q1R28 Independent Shutdown Safety Review Results
(ISRB)
2/13/2025
Q1R28 U0 EDG RMA Discrepancies
03/14/2025
Q1R28 MCC 18-2 RMA Discrepancies
03/16/2025
Corrective Action
Documents
Q1R28 T12/Bus 18 RMA Discrepancies
03/17/2025
Corrective Action
Documents
Resulting from
Inspection
NRC ID: Process and Procedure Enhancement QCOS
0050-06
03/12/2025
Shutdown Safety Management Program
OU-QC-104
Shutdown Safety Management Program Quad Cities Annex
Procedures
QOM-6600-01
Unit 0 Diesel Generator Valve Checklist
U1 HPCI Design Basis Flow Rate/Discharge Pressure
Not Met
05/30/2024
HPCI Min. Acceptable Design Basis Not Met
08/28/2024
Out of Cal. Tolerance but Within Admin Limits
11/25/2024
11/25/2024
U1 HPCI Low Flow Margin
11/26/2024
Did Not Receive Expected Alarm at Alarm Setpoint
2/16/2024
Leak Identified on 1/2 EDGCWP Piping Upstream of
0-3999-139
2/22/2024
Wall Thickness Below Administrative Limit of 0.100"
01/16/2025
NRC/IEMA Questions Regarding EDG Air Line Lubricator
01/22/2025
Corrective Action
Documents
U1 SBO 1B Jacket Water Booster Pump Failed to
Auto-Start
01/29/2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
HPCI FT 1-2358 Miscalibrated
2/07/2025
QCOS 2300-05 Acc. Crit. Not Met Due to Miscalibrated FT
2/18/2025
Q1R28:MOV 1-1001-19A Underthrust
03/12/2025
PSU - Received 902-3 F-14 HPCI Lo Flow and MGU Not at
03/22/2025
OSP-1/2 EDG Vent Fan Started from ALT Feed - QCOS
6600-43
03/25/2025
QCAN 901(2)-3 B-
Core Spray/RHR Fill System Failure
Drawings
4E-1351B, Sheet 1
Schematic Diagram Diesel Generator 1-2 Auxiliaries and
Start Relays
N-513-4 Evaluation for DGCW Line 0-3967-8"-O
Engineering
Changes
ECR 466599
Evaluate HPCI Condition Described in AR 04847782
03/22/2025
HIGH/LP ECCS Discharge Lines Setpoint Error Analysis at
Normal Operations Conditions
000
QC-18-009
SURVEILLANCE TEST INTERVAL (STI) EVALUATION
FORM
Engineering
Evaluations
IM Perform U1 QCIS 1000-11
2/14/2024
IST-QDC-BDOC-
V-19
Inservice Testing Basis Document for RHR-A Loop Cross
Tie Line Isolation
08/16/2024
Miscellaneous
VETIP NIP-302
Oil Fog Lubricator for Compressed Air Service for the EDG
10/1978
Operability
Evaluations
OpEval 643320
Operability Evaluation of the 0-3967-8" line
PCI System Switch Development
05/17/2001
PCI System Atmospheric Control System Outboard
07/11/2014
AR 4827645 AO-2-1601-58 Closed Immediately
Upon Opening
01/06/2025
MA-QC-736-100
Quad Cities Lubrication Guide
QCAN 901(2)-3 B-
Core Spray Discharge Header High/Low Pressure
QCMMS 6600-03
Emergency Diesel Generator Periodic Preventative
Maintenance Inspection
Procedures
QCOS 1600-14
Pressure Suppression System Power Operated Valve IST
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Testing
QCOS 6600-10
1/2 Diesel Vent Fan Auto-Transfer Logic Test
QCOS 6600-43
Unit 1/2 Emergency Diesel Generator Load Test
Leak Identified on 1/2 EDGCWP Piping Upstream of
0-3999-139
01/03/2025
Work Orders
OP 1/2 DG Monthly Load Test (IST)- QCOS 6600-43
2/19/2025
Corrective Action
Documents
Error in QCOP 0201-14
03/10/2025
Provide a Documented Engineering Review of
QCOP 0201-13
Engineering
Changes
Provide a documented Engineering review of
QCOP 0201-14
Temporary Configuration Changes
QCIPM 0200-03
Outage Reactor Water Level Instrumentation Setup
QCOP 0201-13
Reactor Level Upper Wide Range Reference Leg Extension
Use and Control
Procedures
QCOP 0201-14
Reactor Vessel Level Control Using a Local Pressure
Gauge or Transmitter
16B
Corrective Action
Documents
Resulting from
Inspection
OSP NRC ID: FS 1-0262-5A Flexible Conduit is Degraded
03/14/2025
Primary Containment Leakrate Testing Program
Fatigue Management and Work Hour Limits
Procedures
Outage Scope Control
Calibration
Certificate
0050144281
Omega DPG1000 Series, Pressure
07/29/2024
Calibration
Records
Certificate of
Calibration
0011553059
SBLC Liquid Flow Rotameter
07/24/2024
Leak Identified on 1/2 EDGCWP Piping Upstream of
0-3999-139
2/22/2024
Corrective Action
Documents
2B SBLC Pump Trip Following Local Start
01/02/2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2-1601-55 Did Not Meet Stroke Time Per QCOS 6600-14
01/06/2025
Closure of Operations PMTs
01/06/2025
AR 48334556
1-6641-MB2 OOT & Replaced
2/04/2025
U1 EDG TDR Timed Out Outside Criteria
2/05/2025
Missing Labels Inside Panel
2/05/2025
PSU Q1R28 MSIV 1-0203-2A LLRT Exceeded
T.S. Limit < 62.4 scfh
03/10/2025
PSU Q1R28 MSIV 1-0203-2B LLRT Exceeded
T.S. Limit < 62.4 scfh
03/10/2025
PSU Q1R28 MSIV 1-0203-2C LLRT Exceeded
T.S. Limit < 62.4 scfh
03/10/2025
PSU Q1R28 MSIV 1-0203-2D LLRT Exceeded
T.S. Limit < 62.4 scfh
03/10/2025
PSU MSIV 1-0203-1B LLRT Exceeded Alert Limit 21.2 scfh
03/10/2025
PSU Q1R28 LLRT: AO 1-2001-3 Exceeded Required
Action Limit
03/11/2025
PSU Q1R28 1-0203-1C LLRT Leakage Above Tech Spec
Limit
03/12/2025
PSU MSIV 1-0203-1D Exceeded Admin Warning
Limit < 31.2 scfh
03/12/2025
PSU WO Request: 1-0220-59A Pressure Decay Test
03/13/2025
PSU 1-0220-59A Unable to Close during QCOS 3200-04
03/14/2025
OPS-ROD H-3 (30-11) Slow During QCOS 0300-27
03/28/2025
OSP-Rod P-5 (54-19) Slow During QCOS 0300-27
03/28/2025
OPS - Q1R28 RCPB Class 1 1D MSIV 1DPM Leak During
Hydro
03/26/2025
Rebuild 1-02200-62B Trim Set
03/26/2025
OSP - AO 1-2001-3 HAS PACKING LEAKAGE
03/27/2025
OSP - Q1R28 1-2001-3 LLRT Exceeded Alert Limit < 5 scfh
03/26/2025
LL Q1R28 Hydro Packing Team Prep
04/03/2025
AR 5089285089282B SBLC PP C/S Erratic
07/13/2006
QCOS 3200-04
REACTOR FEEDWATER CHECK VALVE 1(2)-220-59AB
CLOSURE TEST
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
Resulting from
Inspection
OSP - NRC ID 6dpm Leak on 1/2 EDG Cooling Line
03/26/2025
Drawings
Schematic Diagram Standby Liquid Control (Unit 2)
08/11/2000
Engineering
Changes
MSIV Stroke Timing Technical Basis
CV-0-59
Condition Monitoring Plan for Check Valve Group 0220-59
03/23/2021
Miscellaneous
Work Request for Cleaning of the Valve Stem 2-1601-55
01/08/2025
DL750 Scopecorder for U1 EDG Relay Testing Incorrectly
Set
2/03/2025
Inservice Testing Program Plan Format and Content
IST-QDC-PLAN
Quad Cities Inservice Testing (IST) Program Plan
01/06/2025
QCEPM 0700-18
Testing and Calibration of Diesel Generator Time Delays
QCOS 0100-05
Main Steam Isolation Valve Local Leak Rate Test
QCOS 0201-08
Reactor Vessel Class 1 and Associated Class 2
System Leak Test
80a
QCOS 0250-014
MSIV Closure Timing
QCOS 0300-27
Control Rod Scram Timing During Vessel System
Leak Test
QCOS 1100-07
SBLC Pump Flow Rate Test
QCOS 1600-08
Containment Isolation Valve Inoperable Outage Report
QCOS 1600-14
Pressure Suppression System Power Operated Valve IST
Testing
QCOS 1600-50
18-inch Primary Containment Vent and Purge Air Operated
Valves Surveillance
QCOS 6600-37
Unit One Div 1 Emergency Diesel Generator Largest Load
Reject Surveillance
QCOS 6600-51
Unit 1 Emergency Diesel Generator Start Failure Logic Test
Procedures
QCOS-0100-07
Feedwater Check Valve Local Leak Rate Test CK 1(2)-
20-58A/B, CK 1(2)-0220-62A/B
CRD SCRAM TIMING DURING RX HYDRO (IST)
03/25/2025
2B SBLC Flow Rate Test
2/14/2023
Work Orders
2B SBLC Flow Rate Test
04/06/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2B SBLC Flow Rate Test
07/05/2024
2B SBLC Flow Rate Test
01/02/2025
Pressure Suppression Valve Timing Test (IST)
01/05/2025
Leak identified on 1/2 EDGCWP Piping Upstream of
0-3999-139
2/27/2024
QDC 2024 NRC Hostile Action Exercise - Lessons Learned
2/27/2024
PUA Issues in Simulator during 2024 NRC Exercise
2/27/2024
1/14/2025 ERO Team 4 Drill Summary
01/15/2025
25 Quad Cities Off Year Exercise: TSC
2/04/2025
25 Quad Cities Off Year Exercise: Drill Dev and Control
2/04/2025
25 Quad Cities Off Year Exercise: Simulator
2/04/2025
Corrective Action
Documents
25 Quad Cities Off Year Exercise: OSC
2/04/2025
On Shift Dose Assessment
Standardized Radiological Emergency Plan
Radiological Emergency Plan for Quad Cities
Addendum 3
Emergency Action Levels
Emergency Classification and Protective Action
Recommendations
ERO Readiness - Performance Indicators Guidance
ER-AA-111-F-06
Quad Cities PAR Flowchart
Procedures
Monthly Data Elements for NRC ROP
Indicator - Emergency Response Organization (ERO)
Drill Participation
QC-01-25-00416
Outboard Main Steam Isolation Valve Overhaul & Seat
Replacement
03/05/2025
QC-01-25-00508
Drywell Nuclear Instrumentation System
03/05/2025
QC-01-25-00541
Drywell Inboard MSIV Overhaul
03/07/2025
QC-01-25-00701
Torus Diving Activities
03/07/2025
ALARA Plans
QC-01-25-00807
TB Building Moisture Separator/Steam Reheater Activities
03/07/2025
Miscellaneous
Alpha Level Assessment
Radiation
Survey #2025-
Bonnet Removal Breach from MSIV-C
03/14/2025
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
236275
Survey #2025-
236341
MSIV-C Supplemental Surey
03/14/2025
Survey #2025-
236373
Torus Desludge Filters
03/15/2025
Survey #2025-
236383
Removal of MSIV-C Internals
03/15/2025
Surveys
Survey #2025-
236763
Torus
03/18/2025
QC-01-25-00508
Drywell Nuclear Instrumentation System
QC-01-25-00541
Drywell Inboard MSIV Overhaul
Radiation Work
Permits (RWPs)
QC-01-25-00701
Torus Diving Activities
0-7504-A SBGT HEPA Filter Access Hatch Gasket
Degraded
08/05/2024
0-7504-B SBGT HEPA Filter Access Hatch Gasket
Degraded
08/05/2024
Corrective Action
Documents
Protected Area SCBA Mako Compressor Needs Revised
11/26/2024
Corrective Action
Documents
Resulting from
Inspection
NRC ID - Clarification Needed for CNP Redon Fit Testing
03/20/2025
e00401530d675c0f
11/19/2024
LAI323977
M7XT SCBA Test Results
11/18/2024
LAI324158
M7XT SCBA Test Results
11/20/2024
MSA FireHawk Air Mask Inspection - Mask MR116 and
Cylinder ACU-126465
03/19/2025
Miscellaneous
S3564.01
Quarterly Service Air and Self-Contained Breathing
Apparatus - Performed 2.25.25 (test results)
2/28/2025
QCIS 7500-01
TIC-3860
Standby Gas Treatment DOP-Freon Test (CM-6.7.1)
(Temporary Change)
10/16/2024
RP-AA-444 TIC-
3858
Controlled Negative Pressure (CNP) Fit Testing
(Temporary Change)
2a
Procedures
Monthly Inspection and Maintenance of MSA FireHawk
Mask Mounted Regulator SCBAs
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Self-Assessments AR 4540872-12-01
Respiratory Protection Self-Assessment
10/31/2023
(LR) SGTS In Place DOP Leak Test of HEPA
Filters-Train B
10/16/2024
Work Orders
(LR) SGTS In Place Charcoal Adsorber Freon Leak
Test - Train B
10/16/2024
Intake Investigation Form
04/02/2024
Intake Investigation Form
03/26/2024
Calculations
Waste Stream Review and Scaling Determination Report
(2023 DAW)
04/04/2024
Corrective Action
Documents
Electric Dosimeter Calibration Mismatch Error in Sentinel
2/24/2024
100518-0
NVLAP Certificate of Accreditation to ISO/IEC 17025:2017
01/01/2025
Multipack 863106
Extremity Dosimeter Results
04/23/2024
Miscellaneous
Multipack 863121
Extremity Dosimeter Results
04/23/2024
Dosimetry Issue, Usage, and Control
Methods for Estimating Internal Exposure from In Vivo and
In Vitro Bioassay Data
Procedures
CFR 61 Program
Radiation
Surveys
L103486-1
Annual 10 CFR 61 DAW Smears (2023)
11/04/2023
Self-Assessments AR 4741597-04
Radiological Hazard, Airborne, and Occupational Dose
Assessment
2/10/2025
QUA-1-1001-20
MIDACALC RESULTS QUA-1-1001-20 (QUA-1)
QUA-2-1001-23A
MIDACALC RESULTS QUA-2-1001-23A (QUA-2)
QUA-2-1001-37A
MIDACALC RESULTS QUA-2-1001-37A (QUA-2)
QUA-2-1001-7A
MIDACALC RESULTS QUA-2-1001-7A (QUA-2)
Calculations
QUA-2-2301-3
MIDACALC RESULTS QUA-2-2301-3 (QUA-2)
NRC CETI: MOV Valve Factors Below Basis Value
11/01/2023
MOV 1-1001-26A Valve Factor Revision and Unacceptable
Margin
11/08/2023
Corrective Action
Documents
NRC NCV 2023-010-04 Non-Conservative Valve Factors
01/19/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
Resulting from
Inspection
NRC ARDI ID: HPCI 10 Minute Mission Time Basis
08/27/2024
HPCI Inputs Validated for LOCA Analysis
Documentation of the Basis for the HPCI 10 Minute Mission
Time
Engineering
Changes
GE-NE-0000-
0034-9313-R0
HPCI System Modification Feasibility Study for Quad Cities
Nuclear Station Units 1 & 2 and Dresden Nuclear Station
Units 2 & 3
Engineering
Evaluations
GE-NE-0000-
0036-4362-01
Small Break LOCA - Dresden and Quad Cities Maximum
Suppression Pool Temperature
During HPCI Operation
05/17/2005
Miscellaneous
NES-MS-06.6
MOV Valve Factors
Procedures
MOV Rising Stem Motor Operated Valve Thrust and Torque
Sizing and Set-Up Window Determination Methodology
Corrective Action
Documents
U2 Turbine Trip and SCRAM
05/23/2024
Miscellaneous
Licensee Event
Report 265/2024-
2-00
Turbine Trip and Automatic Scram due to Digital EHC
Power Supply Intermittent Failure
07/22/2024
Self-Assessments RCR 04776191-11
Root Cause Report for the Unit 2 Scram
08/20/2024