IR 05000254/2024002
| ML24222A627 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 08/12/2024 |
| From: | Robert Ruiz NRC/RGN-III/DORS/RPB1 |
| To: | Rhoades D Constellation Energy Generation |
| References | |
| IR 2024002 | |
| Download: ML24222A627 (1) | |
Text
SUBJECT:
QUAD CITIES NUCLEAR POWER STATION - INTEGRATED INSPECTION REPORT 05000254/2024002 AND 05000265/2024002
Dear David Rhoades:
On June 30, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Quad Cities Nuclear Power Station. On July 16, 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.
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.
August 12, 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:
DPR-29 and DPR-30 Report Numbers:
05000254/2024002 and 05000265/2024002 Enterprise Identifier:
I-2024-002-0066 Licensee:
Constellation Nuclear Facility:
Quad Cities Nuclear Power Station Location:
Cordova, IL Inspection Dates:
April 01, 2024, to June 30, 2024 Inspectors:
J. Cassidy, Senior Health Physicist Z. Coffman, Resident Inspector R. Farmer, Health Physicist M. Garza, Senior Project Engineer T. Hooker, Health Physicist C. Hunt, Senior Resident Inspector T. Okamoto, Project Engineer 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
Failure to Correct a Condition Adverse to Quality Following 10 CFR Part 50, Appendix J,
Containment Local Leak Rate Testing Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000254,05000265/2024002-01 Open/Closed
[H.14] -
Conservative Bias 71152A The inspectors identified a Green finding and associated non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions," when the licensee failed to perform corrective maintenance on reactor core isolation cooling steam supply valve, 2-1301-16, following two consecutive refueling outages with local leak rate tests exceeding the administrative leakage limits established by licensee Technical Specification 5.5.12,
"Primary Containment Leakage Rate Testing Program."
Additional Tracking Items
None.
PLANT STATUS
Unit 1 The unit began the inspection period at full-rated thermal power. On May 19, 2024, the unit performed a downpower to 20 percent for planned maintenance. On May 20, 2024, the unit returned to 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 shut down for refueling outage Q2R27. The unit began power ascension following the Q2R27 outage on April 8, 2024. The unit returned to full power on April 12, 2024. On May 23, 2024, the unit tripped following an unexpected main turbine trip. The unit returned to full power on May 27, 2024. 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 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) (2 Samples)
- (1) Unit 2 125 Vdc charger #2 on May 15, 2024
- (2) 2B core spray partial walkdown during 2A core spray run on June 26, 2024
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (2 Samples)
- (1) Fire Zone 8.2.1.D, Unit 2 turbine building, elevation 558'-6", area below hotwell on April 1, 2024
- (2) Fire Zone 1.1.1.3, Unit 1 reactor building, elevation 623'-0", mezzanine level on June 25, 2024
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the onsite fire brigade performance during an actual event on April 10, 2024.
71111.07A - Heat Exchanger/Sink Performance
Annual Review (IP Section 03.01) (1 Sample)
The inspectors evaluated readiness and performance of:
- (1) ultimate heat sink during dredging activities on May 6, 2024
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 reactor startup following the Q2R27 refueling outage on April 8, 2024.
- (2) The inspectors observed and evaluated licensed operator performance in the control room during reactor startup following the Q2F71 forced maintenance outage on May 25, 2024.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator requalification training on May 16, 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) 4730133, "U1 EDG [emergency diesel generator] Output Breaker Tripped," on April 11, 2024
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (4 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) AR 4768760, "QCOS 1000-43 Relay 10A-K40A Relay Required Adjustment," on April 24, 2024
- (2) AR 4768833, "QCOS 1000-43: Relay 10A-K22A Did Not Energize," on April 24, 2024
- (3) AR 4773448, "1A RB FLR DRN Sump Overflowed While Performing QCOP 1000-27,"
on May 22, 2024
- (4) AR 4776765, "PPC Used for Mitigating Action is Not Functioning Correctly," on May 29, 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) Unit 2, 2A fuel pool cooling water pump temporary modification to support shutdown safety risk during the Q2R27 refueling outage on June 7, 2024
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated refueling outage Q2R27 activities from March 18, 2024, to April 7, 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) (9 Samples)
(1)as-left local leak rate test on main steam isolation valve 2-0203-2A following corrective maintenance on April 2, 2024
- (2) Unit 2 high-pressure coolant injection PMT activities following planned maintenance on April 3, 2024
- (3) Unit 2 reactor core isolation cooling PMT activities following planned maintenance on April 3, 2024 (4)as-left local leak rate test on feedwater check valve 2-0220-62A following corrective maintenance on April 4, 2024
- (5) PMT following leak repairs on standby liquid control valve 2-1101-1 on April 8, 2024
- (6) PMT following corrective work on 2-1001-5A on April 10, 2024
- (7) Unit 2 reactor vessel class 1 and associated class 2 system leak test during refueling outage Q2R27 on April 10, 2024 (8)service level I coating inspections and repair during Q2R27 on May 8, 2024
- (9) Unit 2 high-pressure coolant injection PMT activities following planned maintenance on June 13, 2024
Surveillance Testing (IP Section 03.01) (2 Samples)
- (1) QCOS 6600-47, "Unit Two Division I Emergency Core Cooling System Simulated Automatic Actuation and Diesel Generators Auto-Start Surveillance," on April 4, 2024
- (2) QCOS 6620-10, "SBO [station blackout] DG [diesel generator] 1 Endurance/Margin and Full Load Reject Test," on April 25, 2024
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
- (1) QCOS 1400-01, "1A Core Spray System Flow Rate Test," on April 16,
RADIATION SAFETY
71124.05 - Radiation Monitoring Instrumentation
Walkdowns and Observations (IP Section 03.01) (8 Samples)
The inspectors evaluated the following radiation detection instrumentation during plant walkdowns:
(1)background radiation monitors for Unit 1 and Unit 2 service water (2)effluent radiation monitoring panel in the radiation waste building (3)area radiation monitor by the turbine building fire elevator (4)low and high range area radiation monitors on the refueling floor (5)air monitors in the turbine building, the trackways, and the refuel floor (6)counting instrumentation in the chemistry lab
- (7) Shephard 89 irradiator in the instrument room (8)portable ion chambers stored ready for use in the instrument room
Calibration and Testing Program (IP Section 03.02) (13 Samples)
The inspectors evaluated the calibration and testing of the following radiation detection instruments:
- (1) MGP Instruments, AMP-100, SN #077587
- (2) MGP Instruments, AMP-200, SN #0021989
- (3) MGP Instruments, AMP-200, SN #0020116
- (4) MGP Instruments, Telepole, SN #0027004
- (5) Eberline, RO-20, SN #076455
- (6) Nuclear Research Corporation, ADM-300, SN #0010651
- (7) Thermo Scientific, ESM FH G-L, SN #078027
- (8) Thermo Scientific, ESM FH G-L, SN #078017
- (9) MGP Instruments, RAM GAM 1, SN #0030097
- (10) Thermo Scientific, SAM12, SN #12205
- (11) Eberline, PM7, 12
- (12) Thermo Scientific, PM12, SN #1224
- (13) Mirion, ARGOS-5, SN#1512-222 Effluent Monitoring Calibration and Testing Program Sample (IP Section 03.03) (2 Samples)
The inspectors evaluated the calibration and maintenance of the following radioactive effluent monitoring and measurement instrumentation:
- (1) Unit 1, service water radiation monitor, RE 1-1799-2 (2)main chimney radiation monitors
71124.06 - Radioactive Gaseous and Liquid Effluent Treatment
Walkdowns and Observations (IP Section 03.01) (3 Samples)
The inspectors evaluated the following radioactive effluent systems during walkdowns:
- (1) Unit 1, reactor building ventilation system discharge to main chimney
- (2) Unit 2, reactor building ventilation system discharge to main chimney (3)main chimney gaseous effluent monitor
Sampling and Analysis (IP Section 03.02) (4 Samples)
Inspectors evaluated the following effluent samples, sampling processes and compensatory samples:
- (1) Unit 1, reactor building ventilation system particulate and iodine sampling
- (2) Unit 2, reactor building ventilation system particulate and iodine sampling (3)main chimney noble gas sampling (4)discharge bay
Dose Calculations (IP Section 03.03) (2 Samples)
The inspectors evaluated the following dose calculations:
- (1) Liquid Release Permit L2024112; remediation well RW2
- (2) Gaseous Release Permit G20241617; main chimney noble gas
Abnormal Discharges (IP Section 03.04) (2 Samples)
The inspectors evaluated the following abnormal discharges:
(1)remediation via MW-R-2D2 to discharge bay in 2022 (2)identification of unplanned release from the radwaste building / main chimney area and remediation via QC-RW-1 in
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
MS05: Safety System Functional Failures (SSFFs) Sample (IP Section 02.04)===
- (1) Unit 1 (July 1, 2023, through March 31, 2024)
- (2) Unit 2 (July 1, 2023, through March 31, 2024)
MS06: Emergency AC Power Systems (IP Section 02.05) (2 Samples)
- (1) Unit 1 (July 1, 2023, through March 31, 2024)
- (2) Unit 2 (July 1, 2023, through March 31, 2024)
BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10) (2 Samples)
- (1) Unit 1 (May 1, 2023, through March 31, 2024)
- (2) Unit 2 (May 1, 2023, through March 31, 2024)
BI02: RCS Leak Rate Sample (IP Section 02.11) (2 Samples)
- (1) Unit 1 (October 1, 2023, through March 31, 2024)
- (2) Unit 2 (October 1, 2023, through March 31, 2024)
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
- (1) May 1, 2023, through March 31, 2024 PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample)
- (1) May 1, 2023, through March 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) AR 4763289, "2-1301-16 WO [work order] Was Not Appropriately Scoped into Q2R27," on April 1, 2024
71153 - Follow-up of Events and Notices of Enforcement Discretion Event Follow-up (IP Section 03.01)
- (1) The inspectors evaluated a Unit 2 reactor trip due to a main turbine trip and the licensees response on May 23,
INSPECTION RESULTS
Failure to Correct a Condition Adverse to Quality Following 10 CFR Part 50, Appendix J, Containment Local Leak Rate Testing Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000254,05000265/2024002-01 Open/Closed
[H.14] -
Conservative Bias 71152A The inspectors identified a Green finding and associated non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions," when the licensee failed to perform corrective maintenance on reactor core isolation cooling steam supply valve, 2-1301-16, following two consecutive refueling outages with local leak rate tests exceeding the administrative leakage limits established by licensee Technical Specification 5.5.12, "Primary Containment Leakage Rate Testing Program."
Description:
On March 21, 2022, during the Q2R26 refueling outage, the as-found leakage of the reactor core isolation cooling (RCIC) system steam supply isolation valve, 2-1301-16, failed to meet the administrative leakage limits during local leak rate testing associated with site procedure QCTP 0130-01, Primary Containment Leakrate Testing Program. Specifically, the as-found leakage was determined to be 11.964 standard cubic feet per hour (scfh) with an administrative action limit of less than or equal to 10 scfh. The licensee documented the condition in the corrective action program (CAP) under issue report (AR) 4486439 and performed a technical evaluation per licensee procedure ER-AA-380, Primary Containment Leakrate Testing Program, Section 4.3.6.6, to justify the deferral of corrective maintenance until the next refueling outage scheduled in 2024 (Q2R27).
Subsequently, on March 18, 2024, the licensee performed local leak rate testing during refueling outage Q2R27 on 2-1301-16 and determined that the as-found leakage was 12.66 scfh against the same administrative leakage limit of less than or equal to 10 scfh. The licensee documented the condition in the CAP under AR 4758892. However, contrary to ER-AA-380, the licensee deferred corrective maintenance again until refueling outage Q2R28, scheduled for 2026, because the activities related to repairing 2-1301-16 were inadvertently not scheduled in refueling outage Q2R27.
Following questions from the inspectors regarding the rationale behind the second deferral of corrective maintenance, the licensee performed a causal evaluation under AR 4763289 and determined that the work orders associated with AR 4486439 were not correctly coded during the outage scope control process. Therefore, the corrective maintenance was not included in the outage scope, and thus, resources were not allocated to plan and perform the maintenance during Q2R27.
The primary containment isolation system provides protection against the onset and consequences of accidents involving the gross release of radioactive materials from primary containment by automatically closing the associated primary containment isolation valves in lines penetrating the drywell and suppression chamber during emergency and post-accident periods. Valve 2-1301-16 is a primary containment isolation valve in series with 2-1301-17, both of which provide primary containment isolation of the steam supply to the RCIC turbine from the reactor. The primary containment isolation function of these valves is a safety related function.
The inspectors reviewed the licensees causal evaluation under AR 4763289 and the circumstances surrounding the issue. Licensee procedure ER-AA-380 provides the requirements for the implementation and administration of the licensees 10 CFR 50, Appendix J, program as required under Technical Specification 5.5.12, Primary Containment Leakage Rate Testing Program. Site procedure QCTP 0130-01 describes the site-specific instructions and requirements for Type A, Type B, and Type C leak rate tests as required by the licensees Technical Specifications and supplements the information provided by ER-AA-380. In accordance with 10 CFR 50, Appendix J, local leakage rate tests that measure containment valve leakage rates, such as 2-1301-16, are considered a Type C Test.
Licensee Technical Specification 5.5.12, states, in part, that the primary containment leakage rate testing program will be in accordance with the guidelines contained in NEI 94-01 Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, revision 2-A, dated October 2008. NEI 94-01, Section 10.2, states that administrative limits for leakage rates shall be established, documented, and maintained for each Type C component prior to the performance of local leakage rate testing in accordance with the guidance provided in ANSI/ANS 56.8-2002, sections 6.5 and 6.5.1. The administrative limits assigned to each component should be specified such that they are an indicator of potential valve or penetration degradation.
The NRC approved the use of NEI 94-01, revision 3, with limitations and conditions, in a safety evaluation documented in the Agencywide Documents Access and Management System (ADAMS) under ML12221A202. The inspectors noted that the safety evaluation defines a failure during Type C testing to be when a valve exceeds its administrative leakage limit as defined by NEI 94-01, Section 10.2. Therefore, the inspectors consider the failure of 2-1301-16 to meet its administrative leakage limit, as documented in AR 4486439 and AR 4758892, a condition adverse to quality.
ER-AA-380, revision 16, Section 4.3.6.6, directs the licensee to perform a technical evaluation if actions to bring the leakage down below the administrative limit prior to unit start-up would create unnecessary hardship. If the evaluation concludes that there is no significant safety impact, then the licensee can choose to raise the components leakage limit or defer the maintenance until the next refueling outage. The licensee performed this evaluation following the failure of 2-1301-16 to meet the administrative leakage limit in Q2R26 and elected to defer the corrective maintenance to Q2R27. The inspectors noted that ER-AA-380 only allows for one deferral of corrective maintenance and establishes a backstop of the next refueling outage to complete the maintenance. The inspectors determined that the subsequent deferral of the corrective maintenance on 2-1301-16 following a second failure to meet the administrative leakage requirements during Q2R27 was outside the bounds of ER-AA-380.
Ultimately, the inspectors determined that the combined leakage rates for all penetrations subject to Type B and Type C testing met the overall acceptance criteria for the containment system during Q2R27. However, the inspectors concluded that the licensees failure to perform corrective maintenance on 2-1301-16, within the time frame directed by ER-AA-380, was a failure to correct a condition adverse to quality, and thus a performance deficiency.
Corrective Actions: The licensee has scheduled corrective maintenance on 2-1301-16 for the upcoming Q2R28 refueling outage.
Corrective Action References: AR 4763289, "2-1301-16 WO Was Not Appropriately Scoped into Q2R27," AR 4758892, "Q2R27 LLRT 2-1301-16 Exceeded Admin Req Action Limit"
Performance Assessment:
Performance Deficiency: The inspectors determined that the failure to correct a condition adverse to quality was a performance deficiency. Specifically, the licensee's work management process failed to scope, plan, and execute corrective maintenance to correct a condition adverse to quality associated with primary containment isolation valve, 2-1301-16, following a failure to meet the administrative leakage limits of the licensee's primary containment leakage rate testing program.
Screening: The inspectors determined the performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the licensee's primary containment leakage rate testing program defines an administrative leakage limit as the leakage limit assigned to each Type C tested component as an indication of potential valve degradation. The limit is established at a value low enough to identify and allow early correction to preclude a component failure. A failure, in the context of the licensee's program, is exceeding the administrative limit and not a total failure of the component. In the case of valve 2-1301-16, the administrative leakage limit has already been exceeded for two consecutive refueling outages. NEI 94-01, Section 10.2.3.4, states, in part, that the licensee should perform a causal determination and identify corrective actions that focus on activities that can eliminate the identified cause of the failure with appropriate steps to prevent recurrence. The causal determination and corrective actions should reinforce achieving acceptable performance. Because the licensee chose to place the valve back in service without corrective maintenance, the actions outlined above were not performed by the licensee. Additionally, absent an understanding of the degraded condition, the licensee has no way to predict how further degradation of the valve will manifest itself during the next operating cycle.
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 3, section C, and answered "No" to both 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 licensee chose not to comply with their primary containment leakage rate testing program because resources were not readily available to perform the corrective maintenance without the possibility of extending the refueling outage and there was available margin below the acceptance criteria for the overall allowable containment system leakage with the degraded condition present.
Enforcement:
Violation: 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures be established to ensure conditions adverse to quality, such as failures, deficiencies, deviations, and nonconformances are corrected.
Licensee Technical Specification 5.5.12, Primary Containment Leakage Rate Testing Program, has been established by the licensee to meet the requirements of 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors. Licensee procedure ER-AA-380, Primary Containment Leakrate Testing Program, identifies the requirements for implementation and administration of the Appendix J program. ER-AA-380, Section 4.3.6.6, states if actions required to bring the leakage down below the administrative limit prior to unit start-up would create unnecessary hardship, then perform a technical evaluation. If the evaluation concludes that there is no significant safety impact, then raise the component(s) leakage limit or defer the maintenance until the next refueling outage. Site procedure QCTP 0130-01, Primary Containment Leakrate Testing Program, describes the site-specific instructions and requirements for Type A, Type B, and Type C leak rate tests as required by the licensees Technical Specifications. QCTP 0130-01, Section E.9, establishes an administrative leakage limit of less than or equal to 10 scfh for valves such as 2-1301-16.
Contrary to the above, on March 22, 2024, the licensee failed to correct a condition adverse to quality associated with safety related primary containment isolation valve, 2-1301-16. Specifically, valve 2-1301-16 failed to meet the administrative leakage requirements of procedure QCTP 0130-01 during local leak rate testing on March 21, 2022, which was a condition adverse to quality. Per procedure ER-AA-380, revision 16, Section 4.3.6.6, the licensee opted to defer maintenance until the next refueling outage, which was Q2R27. However, on March 22, 2024, during Q2R27, the licensee chose not to correct the condition adverse to quality associated with 2-1301-16 because the work was inadvertently not scheduled, and thus resources were not immediately available to perform the corrective maintenance, due to a failure of the licensees work management process. As a result, the licensee deferred the corrective maintenance again until Q2R28 and returned the valve to service with a known degraded condition.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On July 16, 2024, the inspectors presented the integrated inspection results to Doug Hild, Site Vice President, and other members of the licensee staff.
- On May 9, 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
Corrective Action
Documents
U2 125VDC Charger Terminal Block
2/27/2023
Drawings
Schematic Diagram 125V DC Battery Charger 2-8300-2
09/09/2015
Procedures
QCOS 1400-10
Core Spray Operability Verification
25 VDC Battery Charger #2 4HR Load Test
01/10/2023
Vent / Backfill Instruments for QCOS 1400-2B CS Pump Flow
03/06/2024
Work Orders
Safety System Manual Valve Position Verification
2/04/2023
Emergency Call for Fire/Smoke
04/10/2024
Smoke Identified in D Heater Bay During QCOS 0010-08
04/10/2024
U-2 D Heater Bay Insulation Reportable Fire Event Critique
04/10/2024
Corrective Action
Documents
Concerns Following Fire Response
04/29/2024
Engineering
Changes
Implement Fire Hose Removal Strategy
2/17/2023
FZ 1.1.1.3
Unit 1 RB 623' Elev. Mezzanine Level
08/2022
Fire Plans
FZ 8.2.1.D
UNIT 2 TB 558-6 ELEV. AREA BELOW HOTWELL
08/2022
Procedures
QCAP 1500-01
Administrative Requirements for Fire Protection
Work Orders
Replace Group B Fire Hoses
05/30/2024
Drawings
B-27A
Intake & Discharge Flume Plan
H
240506 7829
SBES Exelon Post Dredge QA Elevation
05/06/2024
240506 7829
SBES Exelon Post Dredge QA vs Design
05/06/2024
Miscellaneous
240506 7829
SBES Exelon Post Dredge QA vs 2024 Pre-Dredge
05/06/2024
Work Orders
SUMMER READINESS CONTINGENCY FOR INTAKE BAY
DREDGING
05/10/2024
Division 1 EDG Failed to Start due to A3 Speed Pick-up
Amphenol Found Disconnected.
2/03/2011
U1 A Station Blackout Diesel Would Not Shutdown due to
Loose Electrical Connection at Governor
2/11/2014
Corrective Action
Documents
U1 EDG Output Breaker Tripped
01/11/2024
EM U-1 EMERG DIESEL SPEED SENSING CIRCUIT
TESTING & CALIBR.
10/11/2021
Work Orders
(LR) Diesel Generator Periodic Inspection
10/09/2023
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Relay 10A-K40A Found Outside Allowable Value in
Surveillance
10/27/2021
QCOS 1000-43 Relay 10A-K40A Relay Required Adjustment
04/24/2024
QCOS 1000-43: Relay 10A-K22A Did Not Energize
04/23/2024
1A RB FLR DRN Sump Overflowed While Performing QCOP
1000-27
05/11/2024
Corrective Action
Documents
PPC Used for Mitigating Action Is Not Functioning Correctly
05/28/2024
Schematic Diagram RHR System Relay Logic DIV I SH 4
10/27/2005
Wiring Diagram Instrument Rack 2201-59A
11/02/1999
Drawings
M-37
Diagram of RHR Service Water Piping
2/10/1996
MM THERMAL RELIEF VALVE 1-1001-165A TEST AND
REPLACE (IST)
04/29/2015
Work Orders
OP 'A' LOOP LPCI AND RHR NON-OUTAGE LOGIC TEST
10/28/2021
Corrective Action
Documents
NRC ID: Shutdown Safety Near Miss
03/19/2024
Engineering
Changes
Develop Temporary Change for Fuel Pool Cooling Water
Pump (FPCWP, 1(2)-1902-A/B) Bus 18, 19, 28, or 29 Alternate
Power
Procedures
QCOP 1900-39
Unit 2A-2B Fuel Pool Cooling Water Pump Preparation for
Normal or Temporary 13.8 kV Feed
3A
EM INSTALL 2A FPC PUMP TEMP POWER QCOP 1900-39
03/24/2024
Work Orders
CE REMOVE 2A FPC PUMP TEMP POWER IAW QCOP
1900-39
06/07/2024
NRC ID: Shutdown Safety Near Miss
03/19/2024
RP Fatigue Assessment/WHR Waiver
03/31/2024
Corrective Action
Documents
CFR 26 Work Hour Limits Waiver
03/31/2024
QDC-0200-N-
2453
Q2R27 Decay Heat and Related Calculations
2/01/2024
Engineering
Evaluations
QDC-1900-N-
2454
Alternate Decay Heat Removal (ADHR) System Qualification
for Q2R27
2/08/2024
Miscellaneous
Q2R27SSP
Q2R27 Shutdown Safety Plan
Calculations
QDC-0010-M-
0395
Quad Cities Station - Unit 2: Estimation of Non-Insulation
Drywell Debris Source for ECCS Strainer Head Loss
Calculations
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
QDC-1600-M-
0545
Quad Cities Units 1 and 2: ECCS Strainer Head Loss
Estimates
QDC-1600-M-
2013
Quad Cities Unqualified Coating in Containment
U1 SBO Exhaust Fan #1 Drive Fail
2/29/2024
PSU Q2R27 MSIV 2-0203-2A LLRT Exceeded. T.S Limit <78
scfh
03/18/2024
Received Unexpected U-1 SBO DG I/O Cabinet PWR Failure
998
03/21/2024
PSU: Q2R27 FW 2-0220-62A As-Found LLRT Exceeded
Action Limit
03/24/2024
Q2R27: U2 HPCI Front Standard Work Tolerance
03/27/2024
OSP 2-1001-5A Bonnet Bolts Over Torqued
03/27/2024
Q2R27 Service Level I Coating Inspection Results
03/28/2024
SBO 1B Starting Air Compressor RV Constantly Lifting
04/23/2024
LS 2-2365 Level Switch Failed
04/04/2024
Intentionally Abbreviated Maintenance 2-1001-5A
03/30/2024
PSU 2B RHR Start Delay Didn't Meet Criteria (QCOS 6600-47)
04/04/2024
Unexpected Indications during QCOS 2300-0
04/09/2024
HPCI Flow Controller Not Responding as Expected
04/09/2024
U2 HPCI Gland Exhauster 2-2306 Motor Running Currents
High
04/10/2024
RCIC Pump Comprehensive Pump Test
04/10/2024
HPCI Bolting Steam Leak
04/11/2024
EOID Battery Corrosion on U1 SBO Batteries
04/23/2024
LI-6620-111A SBO Oil Sump Level A Failed High
04/25/2024
LI-6620-111B SBO Oil Sump Level B Failed High
04/25/2024
U2 HPCI Will Not Reach High Speed Stop
06/12/2024
NRC ID - Core Thermal Power During HPCI Surveillance
06/14/2024
Corrective Action
Documents
This Evaluation is Intended to Document Quad Cities Review
and Acceptance of the results and Recommendations
Identified in the Underwater Engineering Services, Inc.,
Evaluation Report (Ref. 1) Following Coating Inspections
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Performed during Q2R27.
NUC2400007-L-
CAR-001
COATING ASSESSMENT REPORT
04/03/2024
Miscellaneous
UCC Project No.
2-23256.555
Quad Cities Station - Unit 2: Estimation of Non-Insulation
Drywell Debris Source for ECCS Strainer Head Loss
Calculations
04/01/2024
SAFETY RELATED (SERVICE LEVEL I) PROTECTIVE
ER-AA-335-015-
2013
VT-2 Visual Examination in Accordance with ASME 2013
Edition
QCOS 0100-05
Main Steam Isolation Valve Local Leak Rate Test AO 1(2)-
203-1A/B/C/D, AO 1(2)-0203-2A/B/C/D
QCOS 0201-08
Reactor Vessel Class 1 and Associated Class 2 System Leak
Test
QCOS 1400-01
Core Spray System Flow Rate Test
Procedures
QOS 6600-47
UNIT TWO DIVISION I EMERGENCY CORE COOLING
SYSTEM SIMULATED AUTOMATIC ACTUATION AND
DIESEL GENERATORS AUTO-START SURVEILLANCE
QCOS 0100-07, "Feedwater Check Valve Local Leak Rate
Test CK 2-0020-62A"
03/31/2024
04/10/2024
U1 HPCI FRONT STANDARD AND OIL SYSTEM
OVERHAUL
03/26/2024
OP PMT HPCI SYSTEM TURBINE OVERSPEED TEST
QCOS 2300-07
04/10/2024
(LR) (NEIL) RCIC TURBINE DISASSEMBLY/INSPECTION
04/01/2024
OP LLRT FW Check 2-0220-58A and -62A, QCOS 0100-07
03/25/2024
Reactor Vessel and Class One Piping Leak Test (ISI); NDE
Report 24-VT2-007
04/02/2024
ECCS Simulation Actuation and DG Auto-Start
04/06/2024
OP QCOS 2300-27 HPCI PUMP COMPREHENSIVE TEST
(IST)
04/11/2024
RCIC FLOW RATE TEST COMPREHENSIVE TEST (IST)
04/10/2024
Work Orders
(LR) SBO ENDURANCE/MARGIN TEST
04/25/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
OP QCOS 2300-1 HPCI PUMP LP FLOW RATE
OPERABILITY TEST
04/10/2024
(IST) (NEIL) HPCI Pump Operability
06/12/2024
Perform VT-2 for 2-1101-1 Leak
04/04/2024
05172520
Service Water Mon-Source Check Cycle
01/12/2023
05172521
Service Water Mon-Calibration Cycle
01/12/2023
PM-12 Calibration Data Sheet Attachment 3
03/16/2023
PM-12 Calibration Data Sheet Attachment 3
03/12/2024
ARGOS-5 Calibration Data Sheet
2/15/2024
ARGOS-5 Calibration Data Sheet
01/30/2023
Calibration Data Sheet PM-7 Portal Monitor
2/16/2022
Calibration
Records
Calibration Data Sheet PM-7 Portal Monitor
11/29/2023
Main Chimney SPING Calibration Needs Rescheduled
08/25/2022
Corrective Action
Documents
RY 0-1799-6 Not Installed s Scheduled
06/23/2023
QCOP 1800-01
Operation of ARM Indicator/Trip Units
Operation of Shepherd Model 89 Calibrator
Procedures
Operation and Calibration of the PM-12 Gamma Portal Monitor
4389076
NRC SA: RP Monitoring Instrument Inspection IP 71124.05
2/03/2021
4456144
NRC SA: RP Monitoring Instrument Inspection IP 71124.05
05/26/2022
4469158
22 Radiation Instrumentation Self Assessment
2/14/2022
4469159
Exelon Power Labs Self Assessment
08/30/2022
Self-Assessments
24223
Radiation Monitoring Instrumentation and Radioactive
Gaseous/Liquid Effluent Treatment, NRC Inspection Manuals
71124.05 and.06
04/24/2024
Work Orders
Chimney Flow Rate Indication Calibration
2/16/2024
Corrective Action
Documents
ODCM-Required Missed Samples due to OOS Equipment
10/09/2023
Miscellaneous
Gaseous Release
Permit
Main Chimney Noble Gas
04/25/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
G20241617
Liquid Release
Permit L2024112
Remediation Well RW2
04/25/2024
SVP-23-018
Radioactive Effluent Release Report for 2022
04/28/2023
SVP-24-024
Corrected Radioactive Effluent Release Report for 2022
04/05/2024
SVP-24-029
Radioactive Effluent Release Report for 2023
04/26/2024
Radioactive Effluent Controls Program
CY-QC-110-605
Reactor Building Vent Gaseous & Particulate Sampling
CY-QC-110-606
Main Chimney Gaseous & Particulate Sampling
CY-QC-120-720
Plant Effluent Dose Calculations
CY-QC-120-724
Continuous Liquid Effluent Analysis
Procedures
CY-QC-170-301
Offsite Dose Calculation Manual
Self-Assessments
Radiation Monitoring Instrumentation and Radioactive
Gaseous/Liquid Effluent Treatment, NRC inspection Manuals
71124.05 and.06
04/24/2024
SBGT Removal of Charcoal Adsorber Test Canister
11/13/2023
Work Orders
Verify Sample Data and Complete Testing Criteria
11/21/2023
00181641-01
Perform ODCM Dose Calculations
Various
Reactor Coolant System (RCS) Specific Activity
various
71151
Miscellaneous
various
Q2R26 LLRT on 2-1301-16 Exceeded Admin Required Action
Limit
03/21/2022
Q2R27 LLRT on 2-1301-16 Exceeded Admin Req Action Limit
03/18/2024
2-1301-16 WO Was Not Appropriately Scoped into Q2R27
04/03/2024
Primary Containment Leakrate Testing Program
71152A
Corrective Action
Documents
QCTP 0130-01
Primary Containment Leakrate Testing Program
Corrective Action
Documents
AR 47766191
U2 Turbine Trip and Scram
05/23/2024