IR 05000254/2022004
| ML23039A186 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 02/10/2023 |
| From: | Robert Ruiz NRC/RGN-III/DORS/RPB1 |
| To: | Rhoades D Constellation Energy Generation, Constellation Nuclear |
| References | |
| IR 2022004 | |
| Download: ML23039A186 (1) | |
Text
SUBJECT:
QUAD CITIES NUCLEAR POWER STATION - INTEGRATED INSPECTION REPORT 05000254/2022004 AND 05000265/2022004
Dear David Rhoades:
On December 31, 2022, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Quad Cities Nuclear Power Station. On January 17, 2023, the NRC inspectors discussed the results of this inspection with Brian Wake, 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.
February 9, 2023 This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Robert Ruiz, Chief Reactor Projects Branch 1 Division of Operating Reactor Safety Docket Nos. 05000254 and 05000265 License Nos. DPR-29 and DPR-30
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000254 and 05000265
License Numbers:
Report Numbers:
05000254/2022004 and 05000265/2022004
Enterprise Identifier:
I-2022-004-0054
Licensee:
Constellation Nuclear
Facility:
Quad Cities Nuclear Power Station
Location:
Cordova, IL
Inspection Dates:
October 01, 2022 to December 31, 2022
Inspectors:
Z. Coffman, Resident Inspector
G. Hansen, Senior Emergency Preparedness Inspector
C. Hunt, Senior Resident Inspector
C. Mathews, Illinois Emergency Management Agency
A. Tran, Resident Inspector
Approved By:
Robert Ruiz, Chief
Reactor Projects Branch 1
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Quad Cities Nuclear Power Station, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Licensee Operability Determination Procedure Inconsistent with Regulatory Requirements of 10 CFR 50.55a, "Codes and Standards" Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000254,05000265/2022004-01 Open/Closed
[H.14] -
Conservative Bias 71152S The inspectors identified a very low safety significance (Green) finding and associated non-cited violation of 10 CFR Part 50, appendix B, criterion V, Instructions, Procedures, and Drawings, for the licensee's use of an alternate evaluation methodology for evaluating the structural integrity of a degraded American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code Class 2 and 3 component that is contrary to the regulatory requirements of 10 CFR 50.55a(g). Specifically, the inspectors identified on three separate occasions that the licensee applied an alternate evaluation methodology per corporate procedure OP-AA-108-115, Operability Determinations, to maintain a degraded Code Class 2 or 3 component exhibiting operational leakage in service without the appropriate demonstration of structural integrity per ASME BPV Code, section XI.
Additional Tracking Items
Type Issue Number Title Report Section Status URI 05000254,05000265/20 22001-02 Compliance With 10 CFR 50.55a and the Use of Alternate Operability Evaluations 71152S Closed
PLANT STATUS
Unit 1
The unit began the inspection period at full-rated thermal power. On November 30, 2022, the unit performed an unplanned downpower to approximately 95 percent following a moisture separator drain tank equipment issue. The unit returned to full-rated thermal power on December 1, 2022, where it remained for the rest of the inspection period, with the exception of short-term power reductions for control rod sequence exchanges, testing, and as requested by the transmission system operator.
Unit 2
The unit began the inspection period at full-rated thermal power. On November 4, 2022, the unit was manually tripped due to a 2B feedwater regulating valve failure. The unit returned to full-rated thermal power on November 7, 2022, where it remained for the rest of the inspection period, 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
71111.01 - Adverse Weather Protection
Impending Severe Weather Sample (IP Section 03.02) (1 Sample)
- (1) Imminent winter storm conditions approaching the site on December 21, 2022
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)1B residual heat removal service water (RHRSW) walkdown during 1A RHRSW work window on October 17, 2022 (2)1A standby liquid control (SBLC) during 1B SBLC pump overhaul on October 25, 2022
- (3) Unit 1 emergency diesel generator (EDG) walkdown during 1/2 EDG work on October 31, 2022
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (7 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
- (1) Fire Zone (FZ) 4.0 and 6.3, auxiliary aux electric room and computer room on October 13, 2022
- (2) FZ 8.2.8.E, Unit 1 turbine building main turbine floor on October 26, 2022
- (3) FZ 11.1.3, Unit 1 reactor building, elevation 555'-0", high pressure coolant injection (HPCI) and HPCI access tunnel on November 15, 2022
- (4) FZ 11.1.4, Unit 2 reactor building, elevation 544'-0", HPCI pump room on November 15, 2022
- (5) FZ 9.1, Unit 1 turbine building, elevation 595'-0", diesel generator on November 16, 2022
- (6) FZ 11.2.3, Unit 1 reactor building, elevation 554'-0", 1A core spray and reactor core isolation cooling room on December 05, 2022
- (7) FZ 1.1.1.3, Unit 1 reactor building, elevation 623', mezzanine level on December 29, 2022
71111.11A - Licensed Operator Requalification Program and Licensed Operator Performance
Requalification Examination Results (IP Section 03.03) (1 Sample)
- (1) The inspectors reviewed and evaluated the licensed operator examination failure rates for the requalification annual operating tests administered from October 4, 2022 through November 10, 2022.
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator performance in the control room during the Unit 2 reactor start up from the Q2F26 forced outage on November 6, 2022.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator requalification training in the simulator on October 18, 2022.
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:
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (3 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) E-2 certification meeting and risk management for work week 11/28/2022 on November 28, 2022
- (2) Unit 2 2B feedwater regulating valve (FRV) servo troubleshooting following planned down power on October 24, 2022
- (3) Emergent work following Unit 1 instrument air leak on November 30, 2022
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 4526270, "Unit 1 HPCI Outboard Isolation Valve Packing Leak"
- (2) AR 4530173, "SCRAM Discharge Volume High Level Alarm"
- (3) AR 4533088, "U1 EDG Fuel Oil Transfer Pump did not Fill 1/2A Diesel Fire Pump Day Tank"
- (4) AR 4541355, "'B' Control Room HVAC [heating, ventilation, and air conditioning] Tripped"
71111.19 - Post-Maintenance Testing
Post-Maintenance Test Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated the following post-maintenance testing activities to verify system operability and/or functionality:
[post-maintenance test] Following Overhaul"
- (2) WO 1527382, "U1 SBO [station blackout diesel] PMT After Exciter Limiter Calibration"
71111.22 - Surveillance Testing
The inspectors evaluated the following surveillance testing activities to verify system operability and/or functionality:
Surveillance Tests (other) (IP Section 03.01) (4 Samples)
- (1) QCOS 1100-0, "SBLC Pump a Flow Rate Test," on October 20, 2022
- (2) QCIS 0200-43, "Rx Water Level Low Calibration," on October 24, 2022
- (3) QCOS 5750-02, "'B' CREV [control room emergency ventilation] Surveillance," on November 1, 2022
- (4) QCIS 0200-12, "Reactor Low Pressure (RHR [residual heat removal]-LPCI
[low-pressure coolant injection]) Functional Test," on November 17, 2022
71114.04 - Emergency Action Level and Emergency Plan Changes
Inspection Review (IP Section 02.01-02.03) (1 Sample)
- (1) The inspectors evaluated the following submitted Emergency Action Level and Emergency Plan changes:
Evaluation No. 22-03, Emergency Action Levels for Quad Cities, 02/16/2022
Evaluation No. 22-21, Emergency Action Levels for Quad Cities, 04/26/2022 This evaluation does not constitute NRC approval.
71114.06 - Drill Evaluation
Drill/Training Evolution Observation (IP Section 03.02) (1 Sample)
The inspectors evaluated:
- (1) Drill and Exercise Performance (DEP) simulator, second afternoon scenario on October 11,
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
MS07: High Pressure Injection Systems (IP Section 02.06)===
- (1) Unit 1 (October 1, 2021, through September 30, 2022)
- (2) Unit 2 (October 1, 2021, through September 30, 2022)
MS08: Heat Removal Systems (IP Section 02.07) (2 Samples)
- (1) Unit 1 (October 1, 2021, through September 30, 2022)
- (2) Unit 2 (October 1, 2021, through September 30, 2022)
MS09: Residual Heat Removal Systems (IP Section 02.08) (2 Samples)
- (1) Unit 1 (October 1, 2021, through September 30, 2022)
- (2) Unit 2 (October 1, 2021, through September 30, 2022)
MS10: Cooling Water Support Systems (IP Section 02.09) (2 Samples)
- (1) Unit 1 (October 1, 2021, through September 30, 2022)
- (2) Unit 2 (October 1, 2021, through September 30, 2022)
71153 - Follow Up of Events and Notices of Enforcement Discretion Event Follow up (IP Section 03.01)
- (1) The inspectors evaluated the failure of the Unit 2 2B feedwater regulating valve and licensees response following the insertion of a manual trip of Unit 2 on high reactor water level on November 4,
INSPECTION RESULTS
Licensee Operability Determination Procedure Inconsistent with Regulatory Requirements of 10 CFR 50.55a, "Codes and Standards" Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000254,05000265/2022004-01 Open/Closed
[H.14] -
Conservative Bias 71152S The inspectors identified a very low safety significance (Green) finding and associated non-cited violation of 10 CFR Part 50, appendix B, criterion V, Instructions, Procedures, and Drawings, for the licensee's use of an alternate evaluation methodology for evaluating the structural integrity of a degraded American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code Class 2 and 3 component that is contrary to the regulatory requirements of 10 CFR 50.55a(g). Specifically, the inspectors identified on three separate occasions that the licensee applied an alternate evaluation methodology per corporate procedure OP-AA-108-115, Operability Determinations, to maintain a degraded Code Class 2 or 3 component exhibiting operational leakage in service without the appropriate demonstration of structural integrity per ASME BPV Code, section XI.
Description:
Inspectors documented Unresolved Item (URI) 05000254,05000265/2022001-02, in the NRC Integrated Inspection Report 05000254/2022001 AND 05000265/2022001 following a review of the licensees use of an alternate evaluation methodology for determining the operability of an ASME Code Class 2 or 3 component with operational leakage in accordance with licensee corporate procedure OP-AA-108-115, Operability Determinations, Revision 24. In that review, the inspectors identified three instances where the licensee applied the alternate evaluation method provision of OP-AA-108-115 to address operational leakage:
- AR 4386700, "Pin-hole Leaks Found at Weld for 0-5799-381 B HVAC Train," dated November 27, 2020
- AR 4466003, "Pin-hole Leaks Found at Weld for 0-5799-381 B HVAC Train," dated December 11, 2021
- AR 4467332, "VT-2 Identified Leak During U2 HPCI Run," dated December 17, 2021 OP-AA-108-115, section 4.5.9, Flaw Evaluation, step 3, states, in part:
When a flaw is identified in ASME Class 2 or Class 3 components, then determination of whether the deficient condition results in a Technical Specification (TS) required SSC or a TS required support SSC being inoperable shall be made. The evaluation methodologies used must meet ASME Code, construction code acceptance standards, an NRC-accepted ASME Code Case as listed in Regulatory Guide 1.147, or an NRC approved alternative. If NRC approved code cases or other NRC approved alternate methods are not available to be used, station technical resources should use other alternative evaluation methods as outlined in Appendix A.2 of NEI [Nuclear Energy Institute] 18-03.
The inspectors reviewed NEI 18-03, Operability Determination, appendix A.2, which directs the use of an alternative evaluation method if an NRC approved Code Case or other NRC-approved method is not available. The guidance was incorporated in licensee procedure OP-AA-108-115, Operability Determinations, Revision 24, Attachment 3, and was used to disposition the three instances of operational leakage described above.
For Code Class 2 or 3 components, the licensees Technical Requirements Manual (TRM),section 3.4.a, Structural Integrity, states that if structural integrity of one or more ASME components are not in conformance, the required action is to immediately restore the structural integrity of the affected component to within its limits or isolate the affected component. The TRM bases document under B 3.4.a, Structural Integrity, states, The inspection programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.
The inservice inspection and testing program for AMSE Code Class 1, 2 and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code applicable addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(i).
The licensees TSs define 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).
Operational leakage is leakage through a flaw in the pressure retaining boundary of an ASME BPV Code Class 1, 2, or 3 SSC discovered during the operational life of a nuclear power plant not occurring during an ASME BPV Code required pressure test. The NRC position on operational leakage is that the licensee treats operational leakage the same as if it were discovered during an ASME BPV-Code required pressure test. Specifically, the licensee's TSs and design basis assume that the associated SSCs relied upon to respond to a design basis accident maintain structural integrity during the operation of the plant. Leakage through the pressure retaining boundary of an SSC covered by TSs is a challenge not only to the structural integrity of the SSC per the ASME BPV Code but also a challenge to the safety function of the SSC, and thus the operability of the SSC. The NRC, through 10 CFR 50.55a(g)(4), mandates the use of ASME BPV Code, section XI, throughout the service life of the nuclear power plant. As such, the only NRC-approved methods to evaluate the structural integrity of ASME Code Class 1, 2 and 3 components are section XI of the ASME BPV Code, the applicable Code cases that have been approved by the NRC, or methodologies that have been specifically approved by the NRC through the relief request process pursuant to 10 CFR 50.55a(g)(6)(i).
The inspectors noted that the alternate evaluation methodology directed by NEI 18-03 does not meet ASME BPV Code requirements nor is it approved by the NRC. Therefore, the use of an alternative method in lieu of an approved methodology to disposition operational leakage is contrary to NRC requirements and would require the licensee to submit a proposed alternative in accordance with 10 CFR 50.55a(z), relief through 50.55a(g)(6)(i) for NRC approval, or an assessment that the complete failure of the line in which the leakage occurred would not impede the safety function of the TS required system, or operability of any other TS SSC in the room within which the leak was located, as clarified in Regulatory Information Summary (RIS) 2022-02 discussed below.
On January 6, 2022, the NRC solicited public comments on draft RIS 2021-XX, Operational Leakage, (Agencywide Documents Access and Management System [ADAMS] Accession Number ML21166A122) clarifying the regulatory requirements regarding operational leakage. Those comments were summarized by the NRC staff in a document titled, Analysis of Public Comments on Draft NRC Regulatory Issue Summary 2021-XX, Operational Leakage', (ML22167A003). On May 13, 2022, the Committee to Review Generic Requirements (CRGR) held a public meeting to discuss the industrys concern with the potential backfitting associated with the draft RIS (ML22119A137). Following that public meeting, the CRGR met with the NRC staff to discuss the staffs response to the industrys concerns (ML22220A012). Ultimately, the CRGR determined that the NRC staff had demonstrated a consistent interpretation of the regulations as outlined in the draft RIS and that the proposed RIS was not a new position; therefore, the RIS was not subject to the backfitting provisions of 10 CFR 50.109, Backfitting.
RIS 2022-02, Operational Leakage, was issued on November 3, 2022 (ML22167A002), and highlights previous NRC staff communications regarding operational leakage. The RIS specifically discusses the regulatory requirements of 10 CFR 50.55a(g) applicable to the Quad Cities URI:
Through 10 CFR 50.55a(g), the NRC has required the use of Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, of the ASME BPV Code.Section XI provides methods to address leakage from these SSCs, and the NRC has found that the methods in Section XI provide reasonable assurance of the structural integrity of SSCs and hence the operability of the system.
The regulatory requirement of 10 CFR 50.55a(g) and TS, clarified by this RIS, is that structural integrity must be ensured for an ASME BPV Code Class 1, 2, or 3 component that is required to be operable according to the TS, and the only approved methods for doing so are provided in the ASME BPV Code, as incorporated by reference in 10 CFR 50.55a, Codes and standards. The ASME Code offers multiple methods to meet this requirement, including a series of ASME BPV Code Cases (e.g., N-513, N-705) and Nonmandatory Appendix U, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Piping and Class 2 or 3 Vessels and Tanks, to ASME BPV Code,Section XI. These methods provide specific allowances for leakage from Code Class components to demonstrate that they continue to meet their specified safety function provided that structural integrity is maintained. The NRC staff has reviewed and, in most cases, approved these measures.
In each instance identified by the inspectors above, the licensee documented operational leakage in either a Code Class 2 or 3 component. The licensee determined that there was no applicable NRC approved Code Case to evaluate the through-wall condition, applied an alternate evaluation methodology outside of what is allowed by 10 CFR 50.55a(g) to evaluate the structural integrity of the affected component, determined the affected component and corresponding safety-related system was operable with operational leakage present, and continued to operate the system until Code repairs were made.
After the first degraded condition was identified under AR 4386700 in November 2020, the inspectors discussed their concerns with the use of an alternate evaluation methodology with the licensee. Those concerns were documented under AR 4391077, Use of OP-AA-108-115 for Alternate Methods, dated December 18, 2020. Actions created from AR 4386700 were subsequently put on hold for over a year, in part, as the licensee sought to voice their opposition to the upcoming RIS. During that time, the two additional degraded conditions documented under ARs 4466003 and 4467332 occurred.
Ultimately, the inspectors concluded that the licensee completed Code repairs to the affected components noted above within the allowed TS limiting condition for operation time and therefore there was no ongoing safety concern. However, the inspectors also determined that corporate procedure OP-AA-108-115 was not appropriate for the circumstances as it incorrectly allowed the licensee to disposition operational leakage in Code Class 2 or 3 components in a manner that is contrary to 10 CFR 50.55a(g) requirements. Therefore, the continued use of OP-AA-108-115 would result in subsequent licensee decisions to operate similarly degraded ASME Code Class 2 and 3 components in the future without adequate demonstration of structural integrity and would increase the probability of an inservice failure of a mitigating system necessary to respond to initiating events.
Corrective Actions: Following the issuance of RIS 2022-02, the licensee generated AR 4536886 and developed an operation's standing order restricting the use of the alternate evaluation methodology until OP-AA-108-115 has been updated to align with RIS 2022-02.
Corrective Action References: AR 4391077, "Use of OP-AA-108-115 for Alternate Methods,"
dated December 18, 2020; AR 4503641, "NRC NOV URI 2022-001-02 Alternate Operability Evaluations," dated August 8, 2022; AR 4536886, "NRC RIS 2022-02 Op Leakage inconsistent w/Op Eval Proc," dated November 14, 2022
Performance Assessment:
Performance Deficiency: Title 10 CFR 50, appendix B, criteria V, requires that activities affecting quality shall be prescribed by procedures appropriate to the circumstances and shall be accomplished in accordance with those procedures. Those procedures shall include appropriate quantitative and qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. The inspectors determined that the licensee's allowance for the use of an alternate evaluation methodology, as outlined in licensee corporate procedure OP-AA-108-115, encouraging the disposition of operational leakage in Code Class 2 or 3 components contrary to 10 CFR 50.55a(g) requirements, is a performance deficiency. Specifically, using an evaluation methodology outside of ASME Code requirements, or approved NRC alternatives, does not provide an appropriate demonstration of the structural integrity of a degraded Code Class 2 or 3 component.
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, if left uncorrected, the performance deficiency could result in subsequent licensee decisions to operate future degraded ASME Code Class 2 and 3 components without the appropriate demonstration of structural integrity and could increase the probability of an inservice failure of a mitigating system necessary to respond to an initiating event.
The inspectors reviewed Inspection Manual Chapter (IMC) 0612, "Examples of Minor Issues,"
appendix E, and determined that the more-than-minor discussion under example 3.g.(2) was applicable to this performance deficiency. Specifically, the licensee does not have procedural controls in place to ensure that an appropriate evaluation of the structural integrity of a degraded Code Class 2 and 3 component is done in accordance with the ASME BPV Code, section XI, to ensure that the component can perform its safety function if called upon.
Significance: The inspectors assessed the significance of the finding using IMC 0609, appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors screened the finding in accordance with IMC 0609, appendix A, exhibit 2, and answered "No" to all six screening questions under section A. 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, 10 CFR 50.55a requires that the structural integrity of an ASME Code Class 2 or 3 component be evaluated in accordance with ASME BPV Code, section XI, or other NRC-approved method, regardless of how the leakage is discovered. The use of section XI or other NRC-approved methodologies provides reasonable assurance of structural integrity and often requires actions, such as augmented inspections and increased monitoring, that go beyond the use of engineering judgement to provide that assurance as is the case with ASME Code Case N-513. Despite this, the licensee chose to use an alternate evaluation methodology that was not approved by the NRC and was not subject to those additional actions to provide reasonable assurance of structural integrity.
Enforcement:
Violation: Title 10 CFR Part 50, appendix B, criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures. The licensee established OP-AA-105-118, Operability Determinations, Revision 23, as the implementing procedure for performing operability determinations of SSCs important to safety, an activity affecting quality.
Title 10 CFR 50.55a(g)(4) states:
Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section XI of editions and addenda of the ASME BPV Code that become effective subsequent to editions specified in paragraphs (g)(2) and
- (3) of this section and that are incorporated by reference in paragraph (a)(1)(ii) or
- (iv) of this section for snubber examination and testing of this section, to the extent practical within the limitations of design, geometry, and materials of construction of the components.
2007 Edition ASME section XI, Article IWA-4420 DEFECT REMOVAL REQUIREMENTS, and Article IWA-4421 General Requirements, state:
Defects shall be removed or mitigated in accordance with the following requirements:
- (a) Defect removal by mechanical processing shall be in accordance with IWA-4462
- (b) Defect removal by thermal methods shall in in accordance with IWA-4461
- (d) Defect removal or mitigation by modification shall be in accordance with IWA-4340 Article IWA-4422.1 Defect Evaluation, states:
- (a) A defect is considered removed when it has been reduced to an acceptable size. If the resulting section thickness is less than the minimum required thickness, the component shall be corrected by repair/replacement activities in accordance with this Article.
- (b) Alternatively, the defect removal area and any remaining portion of the defect may be evaluated and the component accepted in accordance with the appropriate flaw evaluation provisions of Section XI, or the design provisions of the Owners Requirements and either the Construction Code or Section III.
Contrary to the above, from June 25, 2020, to November 8, 2022, licensee procedure OP-AA-108-15, Operability Determinations, Revision 23, and subsequent Revision 24, incorrectly allowed the disposition of operational leakage in a Code Class 2 or 3 component using an alternate evaluation methodology, contrary to the requirements of ASME BPV Code, section XI, as incorporated by 10 CFR 50.55a(g)(4), or an NRC-approved alternative pursuant to 10 CFR 50.55a(z) or relief pursuant to 10 CFR 50.55a(g)(6)(i).
As a result, from November 27, 2020, through December 2, 2020, and again on December 11, 2021, through December 13, 2021, the licensee failed to meet requirements set forth in the ASME BPV Code, section XI, for a Code Class 3 component. Specifically, for line 0-57479-2 1/2" connected to the RHRSW/service water supply line to the train 'B' CREV system, with a section thickness less than the minimum required thickness resulting in a through-wall leak, the defect was not removed or mitigated in accordance with article IWA-4421 and not dispositioned in accordance with Article IWA-4422, an NRC-approved alternative pursuant to 10 CFR 50.55a(z), or a relief pursuant to 10 CFR 50.55a(g)(6)(i).
Additionally, contrary to the above, from December 17, 2021, through December 20, 2021, the licensee failed to meet requirements set forth in the ASME BPV Code, section XI, for a Code Class 2 component. Specifically, for line 2-23018B-1/2" that recirculates flow from the discharge of the high-pressure coolant injection booster pump to the in-board booster pump gland seal, with a section thickness less than the minimum required thickness resulting in a through-wall leak, the defect was not removed or mitigated in accordance with Article IWA-4421 and not dispositioned in accordance with Article IWA-4422, an NRC-approved alternative pursuant to 10 CFR 50.55a(z), or a relief pursuant to 10 CFR 50.55a(g)(6)(i).
Enforcement Action: This violation is being treated as a non-cited violation, consistent with section 2.3.2 of the Enforcement Policy.
The disposition of this finding and associated violation closes URI:
05000254,05000265/2022001-02.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
On January 17, 2023, the inspectors presented the integrated inspection results to Brian Wake, Site Vice President, and other members of the licensee staff.
On November 29, 2022, the inspectors presented the emergency action level and emergency plan changes inspection results to Mark Humphrey, Regulatory Assurance Manager, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
OP-AA-108-107-
1001
Station Response to Grid Capacity Conditions
OP-AA-108-111-
1001
Severe Weather and Natural Disaster Guidelines
QCOP-0010-02
Required Cold Weather Inspections
Procedures
Seasonal Readiness
FZ 11.1.3
UNIT 1 RB 554-0 ELEV. HPCI & HPCI ACCESS TUNNEL
08/2022
FZ 11.1.4
UNIT 2 RB 544-0 ELEV. HPCI Pump Room
08/2022
FZ 11.2.3
Unit 1 1A Core Spray and RCIC Room
08/2022
FZ 4.0
SB 595'-0" Elev. Computer Room in Auxiliary Electric Room
08/2022
FZ 6.3
SB 595'-0" Elev. Auxiliary Electric Room
08/2022
Fire Plans
FZ 9.1
Unit 1 TB 595'-0" Elevation Diesel Generator
08/2022
Procedures
TQ-AA-150-F25
LORT Annual Exam Status Report
11/10/2022
Corrective Action
Documents
Trip of 'A' CR HVAC
05/29/2022
Electrical Insulation Service Building Plan and Sections EL
23-0
04/07/2014
Drawings
Control Room HVAC Train 'A' Air Handling Unit 0-5795-361
09/08/2011
Upgrade 'A' Train CR HVAC Air Handling Unit and
Associated Equipment
Relocate the 'A' Train of Control Room HVAC Interlock DP
Switch
Engineering
Changes
ECR 455064
Bypass Train A Control Room HVAC VFD Unit
07/26/2022
(a)(1) Determination - Trip of 'A' CR HVAC
07/12/2022
Engineering
Evaluations
a(1) Action Plan Development and Action Plan - Trip of 'A'
CR HVAC
10/21/2022
Procedures
QCOP 5750-29
Operation of Control Room HVAC 'A' Train Service Water
and Chilled Water Systems
Corrective Action
Documents
Received 901-6 A8 and A11 MSDT 1A and 1B DRN Closed
11/30/2022
Miscellaneous
QC-MISC-045
Technical Specification 3.0.10 Risk Evaluation for Removal
of Watertight/Bulkhead Door 2-0030-249
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
On-Line Risk Management
U1 EDG FOTP Did Not Fill 1/2A DFP Day Tank
10/28/2022
'B' Control Room HVAC Tripped
2/07/2022
Corrective Action
Documents
'B' Control Room HVAC Tripped
2/07/2022
Schematic Diagram Control Room HVAC Equipment
2/21/2001
M-29, Sheet 1
Diagram of Fire Pump Diesel Engine Piping
10/07/2003
Drawings
M-29, Sheet 2
Diagram of Diesel Generator Fuel Oil Piping
2/03/2012
Relocate the 'A' Train of Control Room HVAC Interlock DP
Switch
Engineering
Changes
Relocate 'A' Train of Control Room HVAC Interlock DP
Switch or the B Train Auto-Start Logic
Miscellaneous
Quad Cities UFSAR, Section 6.4.4 - Design Evaluations,
Specifically for the Control Room HVAC System
Reportable Event SAF
QCOP 5750-09
Control Room Ventilation System
QCOS 6600-05
Emergency Diesel Generator Fuel Oil Transfer Pump Flow
Rate Test
QOA 5750-09
Loss of 'A' Train Control Room HVAC
QOA 5750-15
Complete Loss of Control Room HVAC
Procedures
QOA 5750-T01
Control Room Standby HVAC System Local Panel
1/2-9400-105 Annunciator Procedures
Corrective Action
Documents
U1 SBO Engine 'B' Turbo Exh Temp Upscale
04/29/2022
MA-QC-772-302
Nuclear Operational Analysis Department Calibration of SBO
DG Excitation Limiter
QCOS 1100-09
SBLC Pump Post Maintenance Packing Test
Procedures
QCOS 6620-01
SBO DG 1(2) Quarterly Load Test
SBLC PP Overhaul
10/27/2022
'B' SBLC Pump Flow Rate Comprehensive (IST)
10/27/2022
Work Orders
11/03/2022
QCIS 0200-12
Reactor Low Pressure (RHR-LPCI) Functional Test
Procedures
QCIS 0200-43
Unit 2 Division II Low and Low Reactor Water Level
Transmitter Calibration
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
QCOS 1100-07
SBLC Pump Flow Rate Test
QCOS 5750-02
Control Room Emergency Filtration System Test
Corrective Action
Documents
Quad Cities EAL Hot Matrix MG1
04/22/2022
Q 22-03
Emergency Action Levels for Quad Cities
2/16/2022
Miscellaneous
Q 22-21
Emergency Action Levels for Quad Cities
04/26/2022
Addendum 3
Emergency Action Levels for Quad Cities Station
Addendum 3
Emergency Action Levels for Quad Cities Station
Procedures
TQ-AA-113-F004
Emergency Preparedness Training - Emergency Action
Levels
71151
Corrective Action
Documents
IR 4464185 Not Screened as a MSPI Failure
09/20/2022
Unit 2 Scram on Rx High Level
11/04/2022
U-2 'A' RPS Failed to Reset
11/04/2022
Control Rod 46-19 Indicates 08 On Full Core Display
11/04/2022
Unit IRM 17 Failed to Insert on SCRAM
11/04/2022
2A ASD Speed Hold During Runback
11/04/2022
Corrective Action
Documents
Unit 2 ASD NXG Swapped to Backup
11/04/2022