IR 05000244/2019010
| ML20181A101 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 06/29/2020 |
| From: | Mel Gray Engineering Region 1 Branch 1 |
| To: | Bryan Hanson Exelon Generation Co, Exelon Nuclear |
| References | |
| IR 2019010 | |
| Download: ML20181A101 (14) | |
Text
June 29, 2020
SUBJECT:
R. E. GINNA NUCLEAR POWER PLANT - REVISED DESIGN BASES ASSURANCE INSPECTION (TEAMS) REPORT 05000244/2019010
Dear Mr. Hanson:
By letters dated October 16, 2019, Agencywide Documents Access and Management System (ADAMS) accession number ML19291A173, and November 27, 2019 (ML19337A431), Exelon Generation Company, LLC (Exelon), the licensee for Ginna Nuclear Power Plant (Ginna)
contested a Notice of Violation (NOV) that was documented in the U.S. Nuclear Regulatory Commissions (NRC) Design Basis Assurance Inspection Report 05000244/2019010 dated September 18, 2019 (ML19261A083). Specifically, the October 16, 2019 letter requested withdrawal of the Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, NOV 05000244/2019010-02, Failure to Ensure Electrical and Seismic Design Requirements of Motor Control Center C and D, concerning failure to assure the design requirements to prevent the loss of MCC H and J due to a seismic event were in place.
By letters dated November 14, 2019 (ML19322A009), and December 19, 2019 (ML19353B357),
the NRC acknowledged receipt of your letters and informed you that the staff would review your basis for contesting the Green NOV and provide the results of our evaluation by written response. The NRC provided our written response in letter to Exelon dated April 16, 2020 (ML20107F834) which described the results of our review and our plans to modify our records to withdrawal the subject NOV. Accordingly, this inspection report is being revised to withdraw NOV 2019010-02. This letter, its enclosures, 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 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding
Sincerely, X /RA/
Signed by: Melvin K. Gray Mel Gray, Chief Engineering Branch 1 Division of Reactor Safety Docket No. 05000244 License No. DPR-18
Enclosure:
Inspection Report 05000244/2019010
Inspection Report
Docket Number: 05000244 License Number: DPR-18 Report Number: 05000244/2019010 Enterprise Identifier: I-2019-010-0052 Licensee: Exelon Generation Company, LLC Facility: R. E. Ginna Nuclear Power Plant Location: Ontario, New York Inspection Dates: June 9, 2019 to June 29, 2019 Inspectors: K. Mangan, Senior Reactor Inspector P. Boguszewski, Project Engineer M. Orr, Reactor Inspector J. Schoppy, Senior Reactor Inspector S. Kobylarz, NRC Contractor C. Baron, NRC Contractor Approved By: Mel Gray, Chief Engineering Branch 1 Division of Reactor Safety Enclosure 1
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a Design Bases Assurance Inspection (Teams) inspection at R. E. Ginna Nuclear Power Plant 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 Perform Testing as Required by ASME Code Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green None 71111.21M NCV 05000244/2019010-01 Closed The team identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a, Codes and Standards, for Exelons failure to perform required testing set forth in the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) for the reactor vessel boundary check valves located in the charging line (check valve 295 and check valve 9314).
Specifically, Exelon did not scope the isolation check valves credited to transition the ASME Class 1 reactor piping from ASME Class 2 piping into their In-service Testing (IST) program and, as a result, the valves have not been tested in the closed direction.
Additional Tracking Items
None.
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 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
===71111.21M - Design Bases Assurance Inspection (Teams) The inspectors evaluated the following components and listed applicable attributes, permanent modifications, and operating experience:
Design Review - Large Early Release Frequency (LERF) (IP Section 02.02) ===
- (1) PS102A - Containment Spray Pump A
- Material condition and installed configuration (e.g., visual inspection/walkdown)
- Normal, abnormal, and emergency operating procedures
- Consistency among design and licensing bases and other documents/procedures
- System health report, maintenance effectiveness and records, and corrective action history
- Design calculations
- Surveillance testing and recent test results
- System and component level performance monitoring
- Range, accuracy, and setpoint of installed instrumentation
- Equipment protection from fire, flood, and water intrusion or spray
- Heat removal cooling water and ventilation
- Energy sources The team used Appendix B guidance for Valves, Pumps, Instrumentation, Electric Loads, and As-Built System.
Design Review - Risk-Significant/Low Design Margin Components (IP Section 02.02)
(5 Samples)
===71111.21M - Design Bases Assurance Inspection (Teams) The team evaluated the following components, permanent modifications, and operating experience during the weeks of June 10th through June 28th. For the components, the team reviewed the attributes listed in IP 71111.21M, Appendix A, Component Review Attributes, such as those listed below. Specifically, the team evaluated these attributes as per IP 71111.21M, Appendix B, Component Design Review Considerations and IP 71111.21M, Appendix C, Component Walkdown Considerations.
- (1) Turbine-Driven Emergency Feedwater Pump (PAF03)
- Material condition and installed configuration (e.g., visual inspection/walkdown)
- Normal, abnormal, and emergency operating procedures
- Consistency among design and licensing bases and other documents/procedures
- System health report, maintenance effectiveness and records, and corrective action history
- Design calculations
- Surveillance testing and recent test results
- Turbine Overspeed Protection
- Pump Steam Binding Protection
- Station blackout environmental conditions
- Seismic/ High Energy Line Break Protection The team used Appendix B guidance for Pumps, Instrumentation, and As-Built System.
- (2) B Emergency Diesel Generator (Mechanical)
- Material condition and installed configuration (e.g., visual inspection/walkdown)
- Normal, abnormal, and emergency operating procedures
- Consistency among design and licensing bases and other documents/procedures
- System health report, maintenance effectiveness and records, and corrective action history
- Equipment/environmental controls and qualification
- Design calculations
- Surveillance testing and recent test results The team used Appendix B guidance for Valves, Pumps, Instrumentation, and As-Built System.
- (3) 480 Volt Motor Control Center C
- Material condition and configuration (e.g., visual inspection during a walkdown)
- Operating environment
- Consistency between station documentation (e.g. procedures) and vendor specifications
- Maintenance effectiveness
- Corrective maintenance records, and corrective action history
- Breaker short circuit capacity
- Bus loading
- Overcurrent protection and coordination The team used Appendix B guidance for Motor Control Centers and Circuit Breakers and Fuses.
- (4) Main DC Distribution Panel A
- Material condition and configuration (e.g., visual inspection during a walkdown)
- Operating environment
- Consistency between station documentation (e.g. procedures) and vendor specifications
- Maintenance effectiveness
- Corrective maintenance records and corrective action history
- Overcurrent protection and coordination The team used Appendix B guidance for Motor Control Centers and Circuit Breakers and Fuses.
- (5) A Standby Auxiliary Feedwater Discharge Valve (9701A)
- Material condition and installed configuration
- Normal, abnormal, and emergency operating procedures
- Consistency among design and licensing bases and other documents/procedures
- System health report, maintenance effectiveness and records, and corrective action history
- Equipment/environmental controls and qualification
- Design calculations
- Surveillance testing and recent test results The team used Appendix B guidance for Valves.
Modification Review - Permanent Modifications (IP Section 02.03) ===
- (2) ECP-16-000114, Protect SI Accumulators from Potential RV 887 Leakage
- (3) ECP-14-000784, MDAFW (Motor Driven Auxiliary Feedwater) Pump Discharge MOVs 4007 and 4008 Replacement
- (4) ECP-17-000007, Bypass Trip Instrumentation - Power Range
- (5) ECP-16-000583, Modification to Replace Allen Bradley Model 700-RTC Non-Digital Relays with Same Model Digital Relays
- (6) ECP-18-000366, Equivalent Change to allow use of Either Elements of Dual Element RTD TE-410A for Reactor Coolant Loop B Leg Temperature
Review of Operating Experience Issues (IP Section 02.06) (3 Samples)
- (1) NRC Information Notice 84-06: Steam Binding of Auxiliary Feedwater Pumps
- (2) OPEX Evaluation for 10 CFR Part 21 Notification P21-04302015; Allen Bradley Timing Relay Model 700RTC
- (3) NRC Information Notice 18-07, Pump/Turbine Bearing Oil Sight Glass Problems, dated June 13,
INSPECTION RESULTS
Failure to Perform Testing as Required by ASME Code Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green None 71111.21M NCV 05000244/2019010-01 Closed
Introduction:
The team identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a, Codes and Standards, for Exelons failure to perform required testing set forth in the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) for the reactor vessel boundary check valves located in the charging line (check valve 295 and check valve 9314).
Specifically, Exelon did not scope the isolation check valves credited to transition the ASME Class 1 reactor piping from ASME Class 2 piping into their In-service Testing (IST) program and, as a result, the valves have not been tested in the closed direction.
Description:
The team identified that check valves 295 and 9314 are credited as the reactor coolant piping (Class 1 piping) and the charging system piping (Class 2 piping) transition.
The team verified that the transition is delineated in the Ginna In-service Inspection program drawings. The valves are located on the two inch charging line that connects the charging system to one of the reactor coolant system (RCS) cold leg pipes. The team also reviewed the Updated Final Safety Analysis Report (UFSAR) which stated the purpose of the charging line check valves is to prevent loss-of-coolant-accident (LOCA) in the event of a rupture of the two inch charging line.
Based on the valves function to prevent a LOCA in the event of a rupture of the two inch charging line, the team questioned Exelon staff on the type of periodic testing or monitoring the valves received. Exelon staff stated only an external visual inspection of the valves is performed and provided the stations response to Generic Letter 87-06 including an associated letter from the NRC, dated June 28, 1983. The correspondence discussed the effects of pipe breaks on structures, systems, and components inside containment and concluded that periodic testing of the check valves was not required. The team reviewed the correspondence and noted that the correspondence had evaluated the effects of a high energy line break caused by a charging system pipe failure as the basis for justifying not testing and the NRC staff concluded that the check valves were not needed to mitigate the dynamic effects of a charging line pipe break on nearby structures, systems, and components. The team noted that there was not an evaluation of the reactor coolant pressure boundary valves design function to prevent a LOCA in the event of a rupture of the two inch charging line. The team asked if an evaluation of a small break LOCA inside or outside containment resulting from a charging system pipe break had been completed which justified not testing the valve closing function.
The team also asked if Exelon staff had requested relief from the ASME Code requirements as part of their 10 year IST submittal to the NRC related to testing of these check valves. The team noted that 10 CFR 50.55a Section (f), Preservice and In-service Testing Requirements, states, in part, that Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements for preservice and in-service testing. ASME OM Code 2004, Paragraph ISTA-1100, defines those pumps and valves which are required to be tested as those who perform a specific function in
- (1) shutting down a reactor to safe shutdown condition,
- (2) maintaining the safe shutdown condition,
- (3) mitigating the consequences of an accident, or
- (4) providing overpressure protection for a system or portions of a system which perform required function in items (1),
- (2) or
- (3) above.
The team determined that the two check valves in the charging line (CV295 and CV9314) are used to establish the reactor coolant pressure boundary transition to the Class 2 charging system piping and are credited to mitigate the consequences of an accident. Therefore, the valves are subject to the requirement of the IST program.
Exelon staff subsequently identified that check valve 9314 was installed in 1985. The valve was installed in order to move the reactor pressure boundary (Class 1 and Class 2 transition)closer to the reactor vessel. However, when the valve was installed and the Class boundary was moved, Ginna staff did not add the valve to the IST program. As a result, the valve had not been included in the ten-year IST program submittal to the NRC.
The team determined Exelon staff had several opportunities to identify this deficiency. When the reactor coolant pressure boundary was moved and, subsequently, during the preparation of each of the required ten-year IST program submittals most recently completed in September 2009 in accordance with the requirements of the ASME OM Code 2004 Edition.
Corrective Actions: Exelon staff entered the condition into the corrective action program.
Additionally, Exelon staff discovered two other valves that were not appropriately scoped into periodic assessment programs, check valve 9313, the charging line auxiliary spray inlet check valve, and check valve 9315, the charging line inlet check valve to loop B hot leg.
Exelon staff determined there was reasonable assurance the valves were functional because check valve 9314 had been replaced during the most recent outage and a containment check valve in the charging system discharge piping had been tested. Additionally, Exelon determined that valves 9313 and 9315 have adjacent normally closed valves for additional redundancy. The team found these conclusions to be reasonable.
Corrective Action References: Action Request 4260192
Performance Assessment:
Performance Deficiency: Exelon failed to perform required testing set forth in the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) for the check valves in the charging line (CV295 and CV9314) in accordance with 10 CFR 50.55a, Codes and Standards.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone.
Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The team determined this finding is of very low safety significance (Green) because the finding would not have resulted in exceeding the RCS leak rate for a small break LOCA and would not have affected other systems used to mitigate a LOCA resulting in a total loss of their function.
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
Violation: Title 10 CFR 50.55a Section (f), Preservice and Inservice Testing Requirements states, in part, that systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements for preservice and inservice testing (referred to in this paragraph
- (f) collectively as inservice testing) of the ASME BPV Code and ASME OM Code. ASME OM Code 2004, Paragraph ISTA-1100, defines those pumps and valves which are required to be tested as those who perform a specific function in
- (1) shutting down a reactor to safe shutdown condition,
- (2) maintaining the safe shutdown condition,
- (3) mitigating the consequences of an accident, or
- (4) providing overpressure protection for a system or portions of a system which perform required function in items (1),
- (2) or
- (3) above.
ISTC-3522, Category C Check Valves, of the 2014 O&M Code states: Category C check valves shall be exercised as follows:
- (a) During operation at power, each check valve shall be exercised or examined in a manner that verifies obturator travel by using the methods in ISTC-5221. Each check valve exercise test shall include open and close tests.
Contrary to the above, Exelon did not perform testing set forth in the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) for the check valves in the charging line (CV295 and CV9314). Specifically, Exelon did not scope the reactor pressure boundary isolation charging line check valves into their IST program and testing of the valve in the closed direction was not performed.
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 June 28, 2019, the inspectors presented the inspection results to William Carsky, Site Vice President and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.21M Calculations DA-EE-92-011-07 Class 1E Motor Control Center Loading Rev. 8
DA-EE-93-104-07 480V Volt Coordination and Circuit Protection Study Rev. 8
DA-EE-96-005-07 Motor Control Center Coordination Analysis Rev. 14
DA-EE-96-005-07 Motor Control Center Coordination Analysis Rev. 0
Corrective Action 04259227
Documents
Corrective Action 04255767
Documents 04255792
Resulting from 04255912
Inspection 04255937
256109
256217
256246
256549
256585
256618
256647
256670
256680
256836
256850
256876
257012
257013
258503
258945
259000
259035
259429
259430
259531
259702
Inspection Type Designation Description or Title Revision or
Procedure Date
259959
260002
260128
260192
260286
Drawings 03200-0102 AC Power Distribution Panels One-Line Diagram Rev. 33
10904-0164 480 Volt Motor Control Center C Schedule Rev. 29
10904-0165 480 Volt Motor Control Center C Schedule Rev. 35
10904-0166 480 Volt Motor Control Center C Schedule Rev. 22
10904-0175 480 Volt Motor Control Center H Schedule Rev. 20
10904-0176 480 Volt Motor Control Center J & K Schedule Rev. 37
10904-0177 480 Volt Motor Control Center L Schedule Rev. 18
10904-0705 480 Volt Motor Control Center N Schedule Rev. 6
33013-1260 Reactor Coolant PI&D Rev. 27
33013-1265 Chemical and Volume Control System Charging (CVCS) Rev. 12
PI&D
33013-2539 AC Plant Load Distribution One Line Wiring Diagram Rev. 31
Miscellaneous Letter from
- R.W. Kober, RG&E, to
- C. Stahle, NRC, Subject: dated
Periodic Verification of Leak Tight Integrity of Pressure 06/11/1987
Isolation Valves (PIV) (Generic Letter 87-06)
Ginna Nuclear Power Plant Integrated Plant Safety dated August
Assessment Systematic Evaluation Process 1983
Letter from
Subject: IPSAR section 4.13, Effects of Pipe Break on 06/28/1983
Structures, Systems, and Components inside Containment
for the R. E. Ginna Nuclear Power Plant
Letter to John
- F. OLeary, Director, Directorate of Licensing, dated
- U.S. Atomic Energy Commission, from Law Offices of 08/15/72
LeBoeuf, Lamb, Leiby & MacRae
Rochester Gas And Electric Corporation R. E. Ginna dated August
Nuclear Power Plant Unit No. 1, Technical Supplement 1972
Accompanying Application For A Full-Term Operating
License, Docket No. 50-244
Application To Convert Provisional Operating License To dated
Inspection Type Designation Description or Title Revision or
Procedure Date
Full-Term Operating License Or Alternately To Extend The 08/09/72
Termination Date Of The Provisional Operating License,
Docket No. 50-244
VTD-W10120- Instructions for Type W Control Centers February
4469 1968
Procedures ER-AA-321-1002 Inservice Testing Program Plan Format and Context Rev.8
Work Orders C92382153