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| {{Adams | | {{Adams |
| | number = ML18127B698 | | | number = ML21081A152 |
| | issue date = 05/07/2018 | | | issue date = 03/19/2021 |
| | title = NRC Design Bases Assurance Inspection (Teams) Inspection Report 05000454/2018010; 05000455/2018010 (DRS-J.Benjamin) | | | title = Closure of Auxiliary Feedwater System Unresolved Item 05000454/2018010-04 and 05000455/2018010-04 |
| | author name = Jeffers M T | | | author name = Stoedter K |
| | author affiliation = NRC/RGN-III/DRS/EB2 | | | author affiliation = NRC/RGN-III/DRS/EB2 |
| | addressee name = Hanson B C | | | addressee name = Rhoades D |
| | addressee affiliation = Exelon Generation Co, LLC, Exelon Nuclear | | | addressee affiliation = Exelon Generation Co, LLC, Exelon Nuclear |
| | docket = 05000454, 05000455 | | | docket = 05000454, 05000455 |
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| | contact person = | | | contact person = |
| | document report number = IR 2018010 | | | document report number = IR 2018010 |
| | | package number = ML21081A138 |
| | document type = Inspection Report, Letter | | | document type = Inspection Report, Letter |
| | page count = 22 | | | page count = 8 |
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| =Text= | | =Text= |
| {{#Wiki_filter:May 7, 2018 | | {{#Wiki_filter:March 19, 2021 |
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| ==SUBJECT:== | | ==SUBJECT:== |
| BYRON STATION, UNITS 1 AND 2NRC DESIGN BASES ASSURANCE INSPECTION (TEAMS) INSPECTION REPORT 05000454/2018010; 05000455/2018010 | | BYRON STATION - CLOSURE OF AUXILIARY FEEDWATER SYSTEM UNRESOLVED ITEM 05000454/2018010-04 AND 05000455/2018010-04 |
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| ==Dear Mr. Hanson:== | | ==Dear Mr. Rhoades:== |
| On February 8, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed a Triennial Baseline Design Bases Assurance Inspection (Teams) at your Byron Station, Units 1 and 2. On April 12, 2018, the NRC inspectors discussed the results of this inspection with Mr. Kanavos, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report. Based on the results of this inspection, the NRC has identified three issues that were evaluated under the risk significance determination process as having very-low safety significance (Green). The NRC has also determined that three violations are associated with these issues. Because the licensee initiated condition reports to address these issues, these violations are being treated as Non-Cited Violations (NCVs), consistent with Section 2.3.2 of the Enforcement Policy. These NCVs are described in the subject inspection report. If you contest the violations or significance of these NCVs, 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 205550001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement; and the NRC Resident Inspector at the Byron Station. If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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 205550001; with copies to the Regional Administrator, Region III; and the NRC resident inspector at Byron Station. 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, Part Inspections, Exemptions, and Requests | | On December 4, 2020, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Byron Station and discussed the results of this inspection with Mr. K. McGuire and other members of your staff. The results of this inspection are documented in the enclosed report. |
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| | No findings or violations of more than minor significance were identified during this inspection. |
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| | 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. |
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| Sincerely, | | Sincerely, |
| /RA/ Mark Jeffers, Chief Engineering Branch 2 Division of Reactor Safety Docket Nos. 50454; 50455 License Nos. NPF37; NPF66 Enclosure: IR 05000454/2018010; 05000455/2018010 cc: Distribution via LISTSERV | | /RA/ |
| | Karla K. Stoedter, Chief Engineering Branch 2 Division of Reactor Safety Docket Nos. 05000454 and 05000455 License Nos. NPF-37 and NPF-66 |
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| =SUMMARY= | | ===Enclosure:=== |
| The U.S. Nuclear Regulatory Commission (NRC) by conducting a Design Bases Assurance Team Inspection at Byron Station, Units 1 and 2, in program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information. NRC-identified findings, violations, and additional items are summarized in the table below. List of Findings and Violations Failure to Prescribe Motor Driven Auxiliary Feedwater Pump Test Procedures that Accounted for the Allowed Emergency Diesel Generator Frequency Variation Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000454/2018010-01; 05000455/2018010-01 Closed [P.2] Problem Identification and Resolution, Evaluation IP71111.21M The inspectors identified a Green finding and an associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, motor driven auxiliary feedwater pump test procedures that accounted for the allowed emergency diesel generator frequency variation. Specifically, the motor driven auxiliary feedwater pump surveillance procedures would allow a pump with degraded and unacceptable performance to meet the test acceptance criteria based upon the test being performed at nominal frequency and not accounting for potentially lower, allowable, emergency diesel generator frequency. Failure to Periodically Test Instrument Air Check Valves Associated with Air-Operated Containment Isolation Valves Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000454/2018010-02; 05000455/2018010-02 Closed None IP71111.21M The inspectors identified a Green finding and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, valves associated with air-operated containment isolation valves. Specifically, the licensee was not periodically testing the check valves designed to close and maintain sufficient pneumatic pressure in the accumulator tanks installed to closed air-operated containment isolation valves 1(2)RF026 and 1(2)RF027 in response to a containment isolation signal.
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| 3 Failure to Verify the Adequacy of the Air Pressure Regulator Setpoint Value for Containment Isolation Valves 1(2)RF026 Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000454/2018010-03; 05000455/2018010-03 Closed None IP71111.21M The inspectors identified a Green finding and an associated NCV of 10 CFR Part 50, Appendix pressure regulator setpoint value for air-operated containment isolation valves 1(2)RF026. Specifically, these safety-related valves were located inside containment but the licensee did not verify that their air pressure regulator setpoint value was adequate to provide the motive force necessary to close them against containment accident pressure and within their allowable stroke times. Additional Tracking Items Type Issue Number Title Report Section Status URI 05000454/2018010-04; 05000455/2018010-04 Use of 10 CFR 50.54(x) for Unit AFW Cross-Tie IP71111.21M Open 4
| | ==Inspection Report== |
| | Docket Numbers: 05000454 and 05000455 License Numbers: NPF-37 and NPF-66 Report Numbers: 05000454/2020010 and 05000455/2020010 Enterprise Identifier: I-2020-010-0055 Licensee: Exelon Generation Company, LLC Facility: Byron Station Location: Byron, IL Inspection Dates: April 13, 2020 to December 04, 2020 Inspectors: E. Sanchez Santiago, Senior Reactor Inspector Approved By: Karla K. Stoedter, Chief Engineering Branch 2 Division of Reactor Safety Enclosure |
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| =INSPECTION SCOPES= | | =SUMMARY= |
| 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 -Water Reactor Inspection Programactivities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
| | The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a design basis assurance inspection (teams) inspection at Byron 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. |
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| ==REACTOR SAFETY==
| | Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information. |
| 71111.21MDesign Bases Assurance Inspection (Teams) The inspectors selected the components listed below based, primarily, on the system approach. The inspectors evaluated the following components and listed applicable attributes; permanent modifications; operating experience; and previously identified inspection issues: Component (4 Samples)
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| : (1) Unit 1 Motor Driven Auxiliary Feedwater (MDAFW) Pump (1AF01PA) a) Material condition and configuration (i.e., visual inspection during a walkdown) b) Operating procedures c) Protection against flooding d) Protection against seismic events e) Maintenance effectiveness f) Component health, corrective maintenance, and corrective action history g) Consistency between station documentation (e.g., procedures) and vendor specifications h) Pump performance/capability calculations i) Runout and minimum flow j) Submergence (e.g., net positive suction head) k) Water supply availability l) Gas intrusion and hydraulic transients m) Surveillance and in-service testing (IST) procedures, acceptance criteria, and results n) Voltage available at motor during degraded voltage conditions o) Motor circuit breaker protective relay settings and calibration results p) Pump break horsepower load on motor
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| : (2) Unit 1 Diesel-Driven Auxiliary Feedwater Pump (1AF01PB) a) Material condition and configuration (i.e., visual inspection during a walkdown) b) Operating procedures c) Protection against flooding d) Protection against seismic events e) Protection against high energy line breaks f) Maintenance effectiveness g) Component health, corrective maintenance, and corrective action history 5 h) Consistency between station documentation (e.g., procedures) and vendor specifications i) Pump performance/capability calculations j) Runout and minimum flow k) Submergence (e.g., net positive suction head) l) Water supply availability m) Gas intrusion and hydraulic transients n) Room heat up and ventilation calculations o) Fuel oil volume consumption/capacity p) Surveillance and IST procedures, acceptance criteria, and results
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| : (3) Unit 1 Diesel-Driven Auxiliary Feedwater Pump Battery (1AF01EA-A) and Charger (1AF01EA-1) a) Material condition and configuration (i.e., visual inspection during a walkdown) b) Operating procedures c) Maintenance effectiveness d) Component health, corrective maintenance, and corrective action history e) Consistency between station documentation (e.g., procedures) and vendor specifications f) Performance and discharge testing g) Battery sizing h) Duty Cycle i) Float and equalize voltages j) Battery loading k) Voltage drop calculation l) Minimum voltage m) Maximum allowed room temperature during normal operations n) Hydrogen concentration evaluation o) Battery life p) Battery charger sizing q) Cable ampacity r) Protective relays/breakers
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| : (4) 120V Alternating Current Instrument Bus (1IP01J) a) Material condition and configuration (i.e., visual inspection during a walkdown) b) Operating procedures c) Protection against flooding d) Protection against a seismic event e) Maintenance effectiveness f) Component health, corrective maintenance, and corrective action history g) Consistency between station documentation (e.g., procedures) and vendor specifications h) Bus loading and voltage i) Inverter and bus capacity j) Overcurrent protection and coordination k) Inverter and constant voltage transformer overcurrent capability l) Protective device selection and settings m) Inverter operation and alarm response procedure n) Load and Technical Specifications surveillance testing 6 Component Large Early Release Frequency (1 Sample)
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| : (1) Containment Floor Drain Discharge Isolation Valves (1/2RF026/27) a) Material condition and configuration (i.e., visual inspection during a walkdown) b) Operating procedures c) Protection against flooding d) Protection against seismic events e) Maintenance effectiveness f) Component health, corrective maintenance, and corrective action history g) Consistency between station documentation (e.g., procedures) and vendor specifications h) Valve performance evaluations (e.g., weak link) i) Instrument air supply and accumulator design j) Air supply pressure control setpoint k) Leakage test procedure and results l) Surveillance and IST procedures, acceptance criteria, and results m) Control logic design Permanent Modification (5 Samples)
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| : (1) Engineering Change (EC) Resolution (Unit (2)
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| : (3) Replacement of Instrument Power Inverter 111 (1IP05E) and Connection to Division 11 constant voltage transformer (CVT) (1IP01E)
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| : (4) Portion of 1A and 2A MDAFW Pumps Unit 1 and 2 Cross-tie L Final Tie-in to Unit 2 for the Unit 1 and 2 MDAFW Pumps Cross- and
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| : (5) Operating Experience (2 Samples)
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| : (1) U.S. Nuclear Regulatory Commission Information Notice 2005-own and
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| : (2) U.S. Nuclear Regulatory Commission Information Notice 84- Review of Previously Identified Inspection Issues
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| : (1) Non-Cited Violation 05000454/2012007-04; 05000455/2012007- and
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| : (2) Non-Cited Violation 05000454/2015008-08; 05000455/2015008-Proper Direction for Low Level Isolation of the Refueling Water Storage Tank in Emergency Operating Procedures.
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| ==INSPECTION RESULTS== | | ===List of Findings and Violations=== |
| 71111.21MDesign Bases Assurance Inspection (Teams) Failure to Prescribe Motor Driven Auxiliary Feedwater Pump Test Procedures that Accounted for the Allowed Emergency Diesel Generator Frequency Variation Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000454/2018010-01; 05000455/2018010-01 Closed [P.2] Problem Identification and Resolution, Evaluation IP71111.21M The inspectors identified a Green finding and an associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, motor driven auxiliary feedwater (MDAFW) pump test procedures that accounted for the allowed emergency diesel generator (EDG) frequency variation. Specifically, the MDAFW pump surveillance procedures would allow a pump with degraded and unacceptable performance to meet the test acceptance criteria based upon the test being performed at nominal frequency and not accounting for potentially lower, allowable, EDG frequency.
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| Description-requirements that allowed a steady-state EDG frequency variation range of 58.8 to 61.2 hertz. On February 26, 2008, the licensee initiated AR 00741054 to document that this EDG frequency variation was not addressed in calculations. At the time of this inspection, the licensee was still in the process of implementing their long-term corrective actions to restore compliance. However, on February 7, 2018, the inspectors noted that the MDAFW pump surveillance test procedures included an acceptance criteria that did not account for the minimum frequency limit value and the test was conducted at the nominal frequency value. Lower frequency conditions reduce pump performance. The licensee subsequently determined that they should have included an administrative limit on the steady state lower frequency band in the MDAFW procedures while they resolve the problem identified in 2008.
| | No findings or violations of more than minor significance were identified. |
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| Corrective Actions: The licensee was still evaluating its planned corrective actions at the time of the inspection. However, the inspectors determined that the continued non-compliance does not present an immediate safety concern because the licensee reviewed recent MDAFW pump surveillance results and determined that there was sufficient margin to account for the frequency variations observed during recent EDG surveillances. In addition, the licensee initiated an extent of condition and preliminarily determined that similar conditions were applicable to other safety-related rotating equipment. Corrective Action Reference: AR 04101772 8
| | ===Additional Tracking Items=== |
| | Type Issue Number Title Report Section Status URI 05000454, Use of 10 CFR 50.54(x) for 71111.21M Closed 05000455/2018010-04 Unit AFW Cross-Tie |
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| =====Performance Assessment:===== | | =INSPECTION SCOPES= |
| Performance Deficiency: The inspectors determined that the failure to prescribe MDAFW pump test procedures that accounted for the allowed EDG frequency variation was contrary to 10 performance deficiency. Screening: The performance deficiency was more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the MDAFW pump surveillance procedures did not ensure that the pump would be capable of providing its minimum required flow rate under lower EDG frequency conditions. Significance: The finding was evaluated using the Significance Determination Process (SDP) in accordance with Inspection Manual Chapter (IMC) Determination Process for Findings At--low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee evaluated the most recent data and reasonably determined that the rotating equipment driven by the EDGs were operable. Cross-cutting Aspect: The finding had a cross-cutting aspect in the Evaluation component of the Problem Identification and Resolution cross-cutting area, which states that the licensee will thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee was still in the process of implementing their long-term corrective actions to restore compliance for the issue identified in 2008 and did not thoroughly evaluate it to ensure that interim resolutions address the impact of the allowed EDG frequency variation to rotating equipment. (P.2)
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| =====Enforcement:=====
| | 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. |
| quires, in part, that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances. The licensee established 1(2)BOSR 5.5.8.AF.5-s 8 (Unit 1) and 7 (Unit 2), as the implementing procedure for MDAFW pump surveillance testing, an activity affecting quality. Contrary to the above, as of February 7, 2018, the licensee failed to have a procedure for conducting MDAFW pump surveillance testing of a type appropriate to the circumstances. Specifically, procedure 1(2)BOSR 5.5.8.AF.5-1c did not account for the allowed EDG frequency variation, which would allow a pump with unacceptable performance to pass the established acceptance criteria and therefore, go undetected when operating at a lower allowable EDG frequency during an actual event. Disposition: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.
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| 9 Failure to Periodically Test Check Valves Associated with Air-Operated Containment Isolation Valves Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000454/2018010-02; 05000455/2018010-02 Closed None IP71111.21M The inspectors identified a Green finding and an associated NCV of 10 CFR Part 50, Appendix instrument air check valves associated with air-operated containment isolation valves. Specifically, the licensee was not periodically testing the check valves designed to close and maintain sufficient pneumatic pressure in the accumulator tanks installed to close air-operated containment isolation valves 1(2)RF026 and 1(2)RF027 in response to a containment isolation signal.
| | ==INSPECTION RESULTS== |
| | Unresolved Item Use of 10 CFR 50.54(x) for Unit AFW Cross-Tie 71111.21M (Closed) URI 05000454,05000455/2018010-04 |
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| =====Description:===== | | =====Description:===== |
| The containment isolation valves 1(2)RF026 and 1(2)RF027 are air-operated -as-they would close, as required, in response to a containment isolation signal, they were provided with air accumulator tanks to preserve sufficient pneumatic pressure in the event of a loss of instrument air. The air accumulator tanks were isolated from the non-safety related air supply by check valves and these check valves needed to close in order to preserve sufficient pneumatic pressure to close the 1(2)RF026 and 1(2)RF027 valves. However, on January 23, 2018, the inspectors noted that these check valves were not subject to periodic testing in the closed direction. Corrective Actions: As an immediate corrective action, the licensee developed a test instruction, tested the check valves, and determined that they were operable. The proposed plan to restore compliance at the time of the inspection included implementing periodic testing of the instrument air check valves in the closed direction. Corrective Action References: AR 04096766 and AR 04098736
| | In 2008, the licensee added steps to Emergency Operating Procedure (EOP)1/2BFR-H.1, Response to Loss of Secondary Heat Sink, to use the motor-driven auxiliary feedwater (MDAFW) train of a non-accident unit to combat a loss of all feedwater event in the opposite unit by using a recently installed unit cross-tie. The EOPs also directed operators to enter the technical specification limiting condition for operation action statement for the unit donating the MDAFW train because the MDAFW trains were not licensed to be shared between the reactor units. |
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| =====Performance Assessment:=====
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| Performance Deficiency: The inspectors determined that the failure to periodically test the check valves associated with air-operated containment isolation valves 1(2)RF026 and a performance deficiency. Screening: 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 failure to periodically test the check valves associated with air-operated containment isolation valves 1(2)RF026 and 1(2)RF027 would have the potential to allow an inoperable containment isolation condition to go undetected.
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| 10 Significance: The finding was evaluated using the SDP in accordance with IMC 0609 e Determination Process for Findings At--low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components, and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The licensee evaluated equipment condition through testing and reasonably determined that the containment isolation valves were operable. Cross-cutting Aspect: No cross cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
| | In 2011, the resident inspectors documented a Severity Level IV non-cited violation (NCV) of 10 CFR 50.59 after determining the licensees 2008 EOP change resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system or component important to safety. Specifically, the Updated Final Safety Analysis Report described the auxiliary feedwater (AF) trains as non-shared systems. This violation was documented in Inspection Report 05000454/2011004; 05000455/2011004 as NCV 05000454/2011004-02; 05000455/2011004-02, Modification of the Auxiliary Feedwater System Without Prior NRC Approval (REF: ADAMS Accession No. ML113070678). |
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| =====Enforcement:=====
| | As a corrective action to this NCV, the licensee removed the steps in EOP 1/2BFR-H.1, "Response to Loss of Heat Sink," that directed the unit cross-tie to be used and removed credit for the cross-tie in the stations Probabilistic Risk Assessment model. On August 8, 2017, the licensee added direction in EOP 1/2BFR-H.1 to use the AF cross-tie by invoking 10 CFR 50.54(x). The procedure change included a note which stated, If at any time it has been determined that restoration of feed flow to any SG [steam generator] is untimely or may be ineffective in heat sink restoration, then the AF crosstie should be implemented per Step 5 (Page 8). The licensee also added a caution which stated, The AF crosstie should be implemented per Step 5 if other attempts to restore feed flow to the SG(s) will not prevent the initiation of feed and bleed. The DBAI team was concerned the note and caution provided direction to initiate the AF unit cross-tie before bleed and feed, rather than the instruction being provided in the actual procedure step. |
| Violation: Title a test program be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Contrary to the above, as of January 23, 2018, the licensee failed to assure that testing required to demonstrate that the safety-related containment isolation valves 1(2)RF026 and 1(2)RF027 would perform satisfactorily in service was identified and performed in accordance with written test procedures which incorporated the requirements and acceptance limits contained in applicable design documents. Specifically, the licensee was not periodically testing the check valves designed to close and maintain sufficient pneumatic pressure in the accumulator tanks installed to closed air-operated containment isolation valves 1(2)RF026 and 1(2)RF027 in response to a containment isolation signal. Disposition: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Failure to Verify the Adequacy of the Air Pressure Regulator Setpoint Value for Containment Isolation Valves 1(2)RF026 Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000454/2018010-03; 05000455/2018010-03 Closed None IP71111.21M The inspectors identified a Green finding and an associated NCV of 10 CFR Part 50, Appendix pressure regulator setpoint value for air-operated containment isolation valves 1(2)RF026. Specifically, these safety-related valves were located inside containment but the licensee did not verify that their air pressure regulator setpoint value was adequate to provide the motive force necessary to close them against containment accident pressure and within their allowable stroke times.
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| =====Description:=====
| | During the 2018 design basis assurance inspection, the inspectors initiated an unresolved item to document their concerns associated with the licensees actions in response to the 2011 SLIV NCV. Specifically, the inspectors challenged the use of 10 CFR 50.54(x) to implement this change. In addition, the inspectors noted that the licensees 10 CFR 50.59 screening for the procedure change did not include or evaluate the aforementioned note and caution statements to determine whether the 2017 EOP change required prior NRC approval via a license amendment. |
| The air-operated containment isolation valves 1(2)RF026 were located inside containment and relied on air accumulators to supply the motive force necessary to close. Their pressure regulators were set to a value of 60 psig. However, during this inspection, the licensee was unable to retrieve an analysis or vendor document that established the basis for this setpoint value. Upon further review, the inspectors challenged the setpoint value capability to close the valves against containment pressure under accident conditions. As a result, on February 6, 2018, the licensee completed an informal analysis that determined that the 60 psig setpoint value was insufficient to provide the necessary motive force to close the valves against containment accident pressure if the valves degrade to their maximum allowable in-service testing and technical specification stroke time values. Corrective Actions: As an immediate corrective action, the licensee performed an informal analysis and reasonably determined that the valves would close during an accident based on an estimated containment pressure value determined using recent actual valve stroke times. The proposed plan to restore compliance at the time of the inspection included performing a formal calculation to establish an appropriate air regulator setting for valves 1(2)RF026. Corrective Action Reference: AR 04101416
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| =====Performance Assessment:=====
| | Since 2018, the NRC has evaluated the details associated with the above concerns, engaged NRC staff experts, and reviewed licensee procedures and owners group documents related to the technical aspects of the issue to determine whether a performance deficiency or violation occurred. Based on this review, the inspectors did not identify a performance deficiency or violation of requirements. |
| Performance Deficiency: The inspectors determined that the failure to verify the adequacy of the air pressure regulator setpoint value for air-operated containment isolation valves was a performance deficiency. Screening: The performance deficiency was more than minor because it adversely affected the structure, system, or component (SSC) and Barrier Performance attribute of the Barrier Integrity cornerstone objective of ensuring that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to verify the adequacy of the air pressure regulator setpoint value for containment isolation valves 1(2)RF026 does not ensure that the valves would close on a containment isolation signal to protect the public from radionuclide releases caused by accidents or events. Significance: The finding was evaluated using the SDP in accordance with IMC 0609 - finding screened as of very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components, and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The licensee performed an informal analysis and reasonably determined that the containment isolation valves were operable in their current condition (i.e., not degraded to their maximum allowable in-service testing and technical specifications stroke time values). Cross-cutting Aspect: No cross cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
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| =====Enforcement:=====
| | Specifically, regarding the concern the licensee used 10 CFR 50.54(x) to implement a permanent change, the inspector concluded the 2017 EOP revision which added the note, caution and the step for implementing the AF cross-tie by invoking 10 CFR 50.54(x) did not constitute a change as defined in 10 CFR 50.59, Changes, tests and experiments. Title 10 CFR 50.59 defines change as, a modification or addition to, or removal from, the facility or procedure that affects a design function, method of performing or controlling the function, or an evaluation that demonstrated that intended functions will be accomplished. Also, NEI 96-07, Section 1.2.4, Relationship of 10 CFR 50.59 to 10 CFR 50.2 Design Bases, states, 10 CFR 50.59 controls changes to both 10 CFR 50.2 design basis and supporting design information contained in the Updated Final Safety Analysis Report. The AF cross-tie would be implemented during beyond design basis events and would require invoking 10 CFR 50.54(x), therefore the change falls outside the site's licensing and design basis. |
| that the licensee provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Contrary to the above, as of February 6, 2018, the licensee failed to verify the adequacy of design. Specifically, air-operated containment isolation valves 1(2)RF026 were located inside containment but the licensee did not verify that their air pressure regulator setpoint value was adequate to provide the motive force necessary to close the safety-related valves against containment accident pressure and within their allowable stroke times. Disposition: This violation is being treated as a NCV, consistent with Section 2.3.2 of the Enforcement Policy. Unresolved Item Use of 10 CFR 50.54(x) for Unit AFW Cross-Tie 05000454/2018010-004; 050004552018010-004 Opened IP 71111.21M
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| =====Description:=====
| | The inspectors also reviewed the URI concern: |
| In 2008, the licensee added steps to Emergency Operating Procedure (EOP) 1/2BFR-to use the MDAFW train of a non-accident unit to combat a loss of all feedwater event in the opposite unit by using a recently installed unit cross-tie. The EOPs also directed operators to enter the technical specification LCO action statement for the unit donating the MDAFW train because the MDAFW trains were not designed and licensed to be shared between the reactor units. In 2011, the resident inspectors noted that the EOP change resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety previously evaluated in the Updated Final Safety Analysis Report because the Updated Final Safety Analysis Report described the MDAFW trains as non-shared systems. However, the licensee implemented this change without prior NRC approval. As a result, the inspectors documented a Severity Level IV NCV of 10 CFR 50.59 in Inspection Report 05000454/2011004; 05000455/2011004 as NCV 05000454/2011004-02; 05000455/2011004-dwater System Without Prior NRC Accession No. ML 113070678). As corrective actions to this NCV, the licensee removed the steps in the EOPs that directed the unit cross-tie to be used and removed credit for the cross-Risk Assessment model. However, on August 8, 2017, the licensee added direction in EOP 1/2BFR-H.1 to use the Unit Auxillary Feedwater cross-tie by invoking 10 CFR 50.54(x). MDAFW unit cross-tie before bleed and feed.
| | : (1) the added note and caution were the only procedure provisions that provided direction on when to use the MDAFW cross-tie; and (2)those procedure provisions were not included in the 10 CFR 50.59 screening. Per the above discussion, the inspectors concluded the change did not fall under the requirements of 10 CFR 50.59. Regarding the caution statement providing direction, the inspectors noted that the caution includes a should statement which per procedure AD-AA-101-1002, Writers Guide for Procedures and T&RM, denotes a management expectation but does not require or direct a specific action. When the licensee was questioned on the purpose of the caution statement, they indicated it was to let the operators know the AF cross-tie should be implemented after all other options are exhausted and feed and bleed was the only remaining alternative. The licensee also stated that per EOP 1/2BFR-H.1, Step 2, "Check if Bleed and Feed is Required," if the criteria described in the procedure for initiating feed and bleed are met, the licensed operators would be required to implement the bleed and feed actions. |
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| 13 The note stated: is untimely or may be ineffective in heat sink restoration, then the AF crosstie should be ld be implemented per Step 5 if other attempts to restore feed flow to the SG(s) will not prevent the -tie and did not include instructions on when to initiate it. The During this inspection period, the inspectors challenged the use of 10 CFR 50.54(x) to implement this permanent change. In addition, the inspecto10 CFR 50.59 screening for the procedure change did not include in its review the added note and caution statements. Because the added note and caution were the only procedure provisions that provided direction on when to use the MDAFW cross-tie, the 10 CFR 50.59 screening did not review the instructions about when to use the MDAFW cross-tie. As a result, the screening failed to determine that the change may have required a technical specification change and, thus, a license amendment as originally planned. At the end of the inspection, the NRC continued to evaluate if a performance deficiency and or violation occurred. This Unresolved Item will remain open pending the outcome of this ongoing review.
| | There is no allowance to bypass actions related to initiating bleed and feed, once the criteria are met. The licensee also stated the bleed and feed actions would not be delayed to allow implementation of the AF cross-tie. Per the licensee, the added note and caution did not require the use of the cross-tie prior to implementing feed and bleed. Step 5 of procedure 1/2BFR-H.1, "Crosstie Train A AF From Unit 1/2," stated, "Shift Manager has: Determined other heat sink restoration efforts are not available or are untimely; Has implemented 10 CFR 50.54(x); Approved implementation of 1BFSG-3, Alternate Low Pressure Feedwater for AF crosstie." Based on these instructions, the decision to invoke 10 CFR 50.54(x) and implement the cross-tie would be made by the shift manager. The caution also served the purpose of informing the operators that the implementation of the AF cross-tie would require invoking 10 CFR 50.54(x). Per procedure AD-AA-101-1002, one of the purposes of a caution statement is to alert personnel of violations of rules, regulations or work practices. Therefore, the inspectors did not identify a performance deficiency associated with these actions. |
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| ==EXIT MEETINGS AND DEBRIEFS== | | ==EXIT MEETINGS AND DEBRIEFS== |
| The inspectors confirmed that proprietary information was controlled to protect from public disclosure. No proprietary information was documented in this report. On February 23, 2018, the inspectors presented the interim inspection results to Mr. Kanavos, Site Vice President, and other members of the licensee staff. On April 5, 2018, the inspectors presented the final inspection results during the exit meeting to Mr. Kanavos, Site Vice President, and other members of the licensee staff. | | The inspectors verified no proprietary information was retained or documented in this report. |
| | * On December 4, 2020, the inspectors presented the closure of auxiliary feedwater system unresolved item 05000454/2018010-04 and 05000455/2018010-04 inspection results to Mr. K. McGuire, Operations Director and other members of the licensee staff. |
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| =DOCUMENTS REVIEWED= | | =DOCUMENTS REVIEWED= |
| 71111.21MDesign Bases Assurance Inspection (Teams) - Data File; ELMS-AC Block Motor Start Reports; Revision 1A - 19-AN-28; Second-Level and Third-Level Undervoltage Relay Setpoint; Revision 1C - PSA-B-98-05; Analysis of AFWS Pump Suction Transients using SX Water Supply for Byron and Braidwood Stations; Revision 0A - BYR01-086; Motor Operated Valves Actuator Terminal Voltage and Thermal Overload Sizing Calculation Auxiliary Feedwater System; Revision 1A - BAR 2-21-C7; Alarm: BUS 242 Overload or Voltage Low; Revision 11 - BAR 2-22-C7; Alarm: BUS 242 Overload or Voltage Low; Revision 11 - BYR03-171; Verification of Diesel Driven AF Pump Room Ventilation Requirements and Hydrogen Concentration Evaluation; Revision 1 - BYR97-193; Battery Duty Cycle and Sizing for the Byron Diesel Driven Auxiliary Feedwater Pumps and Byron Diesel Driven Essential Service Water Makeup Pumps; Revision 3 - BYR2000-136; Voltage Drop Calculation for 4160V Switchgear Breaker Control Circuits; Revision 3 - NED-I-EIC-0186; Auxiliary Feedwater Pump Suction Pressure Setpoint Error Analysis; Revision 5
| | |
| - CN-CA22-009; RSAC ATWS Analysis for BY1C22; 11/07/2016 - BYRO-043; Documentation of Adequate NPSHa for AF pumps from CSTs; Revision 01 - SX1-89; Available NPSHa for AF pumps from SX System; Revision 1 - PSA-97-18; Byron/Braidwood AFW Flow for AF005A-H Modification with Cross-Tie Effect; Revision 5B - BYR03-095; AFW Strong Pump Weak Pump Interaction on Recirculation Flow; 05/17/2004 - 4391/19AN-3; Protective Relay Setpoints for 4.16 kV ESF Switchgear; Revision 16 - 19-AQ-63; Division Specific Degraded Voltage Analysis; Revision 7 - 19-AQ-67; Voltage Drop on 120V Instrument Inverters; Revision 6 - 19-AQ-66; Instrument Bus Inverter Loading; Revision 4 - BYR13-234/ BRW-13-0216-M; Auxiliary FW Pump Room Temperature Analysis During an ELAP Event; Revision 1 - CQD-200105; Allowable Valve Acceleration of XOMOX Valves; Revision 0 - CQD-220159; Seismic Qualification of Valves Tufline Division XOMOX; Revision 0 - 3C8-1281-001; Auxiliary Building Flood Level Calculations; Revision 9 - BYR04-051/BRW-04-0038-M; Re-Analysis of LOCA Using Alternate Source Terms; Revision 5 - AF-SUM; AF Summary for Pump Performance; Revision 1 - BYR03-095; Auxiliary Feedwater Strong Pump/ Weak Pump Interaction on Recirculation Flow; Revision 0 - BYR04-043; Documentation of Adequate NPSHA for AFW Pumps When Supplied From CSTs; Revision 1 - BRW-10-0146-M/BYR10-103; W Diesel Driven Pump Fuel Consumption and Day Tank Requirements; Revision 3 - BOP SX-1; Essential Service Water Pump Startup; Revision 31 - BOP CC-1; Component Cooling Water System Startup; Revision 12 - 1BEP 0; Reactor Trip or Safety Injection; Revision 301 - 1BEP 1; Loss of Reactor or Secondary Coolant; Revision 300 - 1BEP 2; Faulted Steam Generator Isolation; Revision 300 - 1BEP 3; Steam Generator Tube Rupture; Revision 301 - 1BEP ES-0.1; Reactor Trip Response; Revision 300 - 1BEP ES-1.2; Post LOCA Cooldown and Depressurization; Revision 300 - 1BFR-S.1; ATWS; Revision 301 - 1BCA 0.0; Loss of All AC Power Unit 1; Revision 300 - BD-FR-H.1; Response to Loss of Secondary Heat Sink; Revision 301 - AD-AA-101-1002; - AD-BY-101-1003; EOP Maintenance Program Guideline; Revision 1 - 1BOA ELEC-5; Local Emergency Control of Safe Shutdown Equipment; Revision 106 - 1BOA SEC-7; Auxiliary Feedwater Check Valve Leakage; Revision 104 - 1BOA SEC-8; Steam Generator Tube Leak Unit 1; Revision 111 - 1BOA ELEC-2; Loss of Instrument Bus Unit 1; Revision 109 - 1BOA ELEC-4; Loss of Offsite Power Unit 1; Revision 113 - 1BOA ELEC-7; Loss of Annunciators Unit 1; Revision 3 - 1BOA PRI-13; Recovery from Inadvertent Phase A Containment Isolation; Revision 104 - BAR 1-18-D3; AMS Initiated; Revision 4 - BAR 2-3-E6; AF System Local Control Loss of Control Power; Revision 2 - BAR 2-22-C7; Bus 242 Overload or Volt Low; Revision 1 - BAR 1-21-C7; Bus 141 Overload or Volt Low; Revision 12 - BAR 1-22-C7; Bus 142 Overload or Volt Low; Revision 12 - BAR 1-3-A2; AF Pump Low Suction Pressure; Revision 1 - 1BFR-H.1; Response to Loss of Secondary Heat Sink Unit 1; Revision 201 - 1BFR-H.1; Response to Loss of Secondary Heat Sink Unit 1; Revision 202
| | Inspection Type Designation Description or Title Revision or |
| - 1BFR-H.1; Response to Loss of Secondary Heat Sink Unit 1; Revision 203 - 1BFR-H.1; Response to Loss of Secondary Heat Sink Unit 1; Revision 300 - BOP AF-4; Diesel Driven Auxiliary Feedwater System; Revision 1 - BOP AF-M1; Auxiliary Feedwater System Mechanical Lineup; Revision 19 - BOP CD-M1; Condensate System Valve Lineup; Revision 26 - BOP DO-13; Filling the Unit 1 Diesel Auxiliary Feedwater Pump Day Tank; Revision 15 - BOP DO-16; Filling the Unit 2 Diesel Auxiliary Feedwater Pump Day Tank; Revision 18 - OP-AA-102-104; Unit 1/2 Standing Order for Auxiliary Feedwater Unit Crosstie Valves and 1/2BFR H.1; 09/29/2011 - 1BDSG-3; Alternate Low Pressure Feedwater Unit 1; Revision 2 - OP-AA-102-106; Operator Response Time Program; Revision 4 - OP-BY-102-106; Operator Response Time Program at Byron Station; Revision 11 - MA-AP-773-531; Unit 1 4KV Bus 141 Cubicle Relay Routine; Revision 2 - BOP IP-1; Instrument Bus Inverter Startup; Revision 22 - BOP IP-2; Transferring an Instrument Bus from the Inverter to the Constant Voltage Transformer; Revision 25 - BOP IP-5; Removing AC Input from an Instrument Bus Inverter; Revision 6 - BOP IP-6; Restoring AC Input to an Instrument Bus Inverter; Revision 8 - BOP IP-7; Operation of the Constant Voltage Transformer with the Instrument Inverter Supplying the Respective Bus; Revision 3 - BOP IP-8; Startup of the Constant Voltage Transformer to Power the Instrument Bua with the Associated Inverter unavailable; Revision 3 - BAR 1-4-A5; Alarm No. 1-4-A5 Bus 111 Inverter Trouble; Revision 2 - 1BOSR IP-R1; Instructions to Cycle Instrument Bus 111 Distribution Panel Molded Circuit Breakers; Revision 3 - 1BOSR SX-M1; 1A AFW Pump SX Suction Line Monthly Flushing Surveillance; Revision 8 - 2BOSR SX-M1; 2A AFW Pump SX Suction Line Monthly Flushing Surveillance; Revision 11 - 1BOSR 0.5-3.AF.1-1; ASME Surveillance Requirements for the Train A Auxiliary Feedwater SX Supply Valves; Revision 21 - 2BOSR 0.5-3.AF.1-2; ASME Surveillance Requirements for the B Train Auxiliary Feedwater SX Supply Valves; Revision 19 - 2BOSR 5.5.8.AF.5-1c; Comprehensive Inservice Testing (IST) Requirements for the Motor Driven Auxiliary Feedwater Pump 2AF01PA; Revision 7 - 1BOSR 5.5.8.AF.5-1c; Comprehensive Inservice Testing (IST) Requirements for the Motor Driven Auxiliary Feedwater Pump 1AF01PA; Revision 8 - 1BOSR 5.5.8.AF.5-2c; Comprehensive Inservice Testing (IST) Requirements for the Diesel Driven Auxiliary Feedwater Pump 1AF01PB; Revision 14 - 2BOSR 5.5.8.AF.5-2c; Comprehensive Inservice Testing (IST) Requirements for the Diesel Driven Auxiliary Feedwater Pump 2AF01PB; Revision 13 - ER-AA-410-1000; Air Operated Valve Categorization; Revision 3 - 2BOSR 7.5.5-1; Train A Auxiliary Valve Emergency Actuation Signal Verification Test; Revision 16 - 2BOSR 7.5.5-2; Train B Auxiliary Valve Emergency Actuation Signal Verification Test; Revision 16 - BOP DG-11; Diesel Generator Startup; Revision 26 - 1BOSR 8.1.11- - 1BOSR 8.1.14-1; 1A Diesel Generator 24 Hour Endurance Run And Hot Restart Test 18 Month; Revision 21 - 1BOSR 8.1.2-1; 1A Diesel Generator Operability Surveillance; Revision 31 - 1BOSR 8.1.8-1; DG 1A Crosstie to Bus 241 SAT 242-1 and Crosstie to Bus 141 18 Month Surveillance; Revision 15
| | Procedure Date |
| - BOP AP-32; Synchronizing a SAT to a Bus Being Fed by a DG; Revision 4 - 1BEP-1; Loss of Reactor or Secondary Coolant; Revision 300 - 1BOA ESP-1; 4kv ESF/Non-ESF Crosstie; Revision 0 - OBOA PRI-8; Auxiliary Building Flooding; Revision 4 - 1BCA-1.2; LOCA Outside Containment; Revision 300 - 1BEP ES-1.3; Transfer to Cold Leg Recirculation; Revision 301 - BAR 0PL02J-2-B2; Containment Floor Drain Sump2 Level High-High; Revision 52 - AR 00892610; 2009 CDBI Issue Degraded Voltage 5 Minute Timer; 03/13/2009 - AR 01071667; Non-Conservative Degraded Voltage Time Delay; 05/20/2010 - AR 01147052; Degraded Voltage Hit Team Needed; 12/01/2010 - AR 01202766; NRC Question on Auxiliary Feedwater Suction Design Basis; 04/14/2011 - AR 01249731; NRC 2Q2011 NCV-AF Piping Subject to Voiding after Seismic Event; 08/10/2011 - AR 01378533; CDBI Unsupportable Assumption in the Aux Bldg Flood Calc; 05/15/2012 - AR 01377545; NRC CDBI Change Required to Eliminate UFSAR Discrepancy; 06/13/2012 - AR 01398431; NRC CDBI Green NCV-Aux Bldg Flooding not Adequately Addressed; 06/15/2012 - AR 01460439; Part 21 ENS 48650 Anti-Rotation Pin Failure in 10-Inch A; 01/09/2013 - AR 01462853; Part 21 NRC 2013-02-00 Part 21 Anti-Rotation Pin Failure; 01/15/2013 - AR 01483685; Part 21, Flowserve Wedge Pin Failure of an Anchor/Darling; 03/05/2013 - AR 02588343; 1B AF PP Trip during PMT; 11/17/2015 - AR 04028825; 2AF024 Limit Switch Does not Fully Engage; 07/05/2017 - AR 0242884; 1B DG Incomplete Sequence Trip Repeat Occurrence; 12/22/2014 - AR 04064929; DBAI FASA - Key Calc Review AF-SUM; 10/19/2017 - AR 04064912; DBAI FASA - Key Calc Review BYR12-070; 10/19/2017 - AR 02490023; Oil Analysis Trend For 1A AUX FW PP (Non-Emergent); 04/23/2015 - AR 04008372; Oil Leak on Outboard Bearing for 2AF01PA-M; 05/09/2017 - AR 02728306; Outboard Dust Cap/Plug Is Wrong Size; 10/14/2016 - AR 01398434; NRC CDBI Leak Detection for ECCS Flow Path Lacking; 08/08/2012 - AR 02454767; NOS ID No CA to Correct an NRC NCV; 02/18/2015 - AR 01378257; CDBI Question About ECCS Leakage; 06/15/2012 - AR 00160059; DDAF Pump Speed Affect On Maximum Pressure; 05/22/2003 - AR 00741054; EDG Frequency Variation Not Addressed on Calcs; 02/26/2008 - M-1280, Sheet 1; DOST Room Ventilation; Revision U - 6E-1-4030AF12; Auxiliary Feedwater Pump 1B (Diesel Driven) Engine Startup Panel 1AF01J; Revision AH - 6E-1-4030AF13; Auxiliary Feedwater Pump 1B Startup Panel 1AF01J Annunciator; Revision N - 6E-1-4469F; Auxiliary Feedwater Pump 1B Miscellaneous Devices 1AF01PB; Revision Q - 6E-1-4685B; 480V Auxiliary Building ESF MCC 132X3 Section B (1AP24E) 120/208V Distribution Panel; Revision J - 6E-1-4002E; Single Line Diagram 120V AC ESF Instrument Inverter Bus 111 & 113; Revision L - 6E-1-4030IP01, Sheet No. 1; Schematic Diagram 10 KVA Inverter for Instrument Bus 11 (1IP05E) Part 1; Revision A - 6E-1-4030IP01, Sheet No. 2; Schematic Diagram 10 KVA Inverter for Instrument Bus 11 (1IP05E) Part 2; Revision A - 6E-1-4030IP01, Sheet 3; Schematic Diagram 10 KVA Inverter for Instrument Bus 11 (1IP05E) Part 3; Revision A - 6E-1-4030IP01, Sheet 4; Schematic Diagram 10 KVA Inverter for Instrument Bus 11 (1IP05E) Part 4; Revision A
| | 71111.21M Engineering 362168 Installation of the Final Phase of the Motor-Driven Auxiliary 0 |
| - 6E-1-4012A; Key Diagram 120V AC Instrument Bus 111 (1IP01J) ESF Div. 11 Channel 1; Revision Y - 6E-1-4006A; Key Diagram 4160V ESF Swgr Bus 141 (1AP05E); Revision I - 6E-1-4030AF01; Schematic Diagram Auxiliary Feedwater Pump 1A (1AF01PA); Revision AE - 6E-1-4030EF01; Schematic Diagram ESF Sequencing and Actuation Cabinet 1PA13J; Revision W - 6E-1-4030EF11; Schematic Diagram Reactor Protection System Output Relays Development - 6E-1-4246A; Wiring Diagram 120 VAC Instr Bus Distr Pnls 111 & 113 (1IP01J & 1IP03J) (1IP01JA & 1IP03JA); Revision V - EC 362146; Replacement of Instrument Power Inverter 111 (1IP05E) and Connection to Division 11 CVT (1IP01E); Revision 0 - 6E-15-052; 50.59 Screening Replacement of ESF Instrument Power (IP) Inverters and Connection to Constant Voltage Transformers (CVT) Revision 0; Revision 0 - 6G-15-005; 50.59 Evaluation Replacement of ESF Instrument Power (IP) Inverters and Connection to Constant Voltage Transformers (CVT) Revision 0; Revision 0 - 6E-01-0143; 50.59 Screening EC 79089, DCP 9900303; Revision 0 - M-42, Sheet 3; Diagram of Essential Service Water; Revision BE - NP5128- XACT 7075 FS#2 Act. R (Fail Close); Revision 1 - M-37; Diagram of Auxiliary Feedwater; Revision BF - M-50, Sheet 3; Diagram of Diesel Fuel Oil; Revision BA - M-122; Diagram of Auxiliary Feedwater; Revision BE - M-126, Sheet 1; Diagram of Essential Service Water; Revision BG - M-130, Sheet 2; Diagram of Diesel Oil And Fuel Oil Supply; Revision AX - M-48, Sheet 6A; Diagram of Waste Disposal Auxiliary Building Floor Drains; Revision AQ - M-48, Sheet 6B; Diagram of Waste Disposal Auxiliary Building Floor Drains; Revision AV - M-55, Sheet 4; Diagram of Instrument Air Aux. Bldg., Containment, and Fuel Handling Building; Revision S - WO 04669849; 1B Diesel Driven Auxiliary Feedwater Pump Bank A Battery B Quarterly Surveillance; 10/27/2017 - WO 04669851; 1B Diesel Driven Auxiliary Feedwater Pump Bank A Battery A Quarterly Surveillance; 10/27/2017 - WO 04669850; 1B Diesel Driven Auxiliary Feedwater Pump Bank B Battery A Quarterly Surveillance; 10/27/2017 - WO 04669851; 1B Diesel Driven Auxiliary Feedwater Pump Bank B Battery B Quarterly Surveillance; 10/27/2017 - WO 04640238; 2B Diesel Driven AF Pump Monthly Surveillance; 06/16/2017 - WO 01730057; 1AF01PA Comprehensive IST; 07/22/2015 - WO 01871116; Low Power Physics Testing Program; 12/19/2016 - WO 01730057; 1BOSR 5.5.8.AF.5-1C Comprehensive IST for MDAFWP 1AF01PA; 09/12/2015 - WO 01952482; 1BOSR 5.5.8.AF.5-1B Group B IST for MDAFWP 1AF01PA; 12/12/2016 - WO 01817191-50; Verify Static Switch Sensing Circuits Initiate a Transfer from Normal to Bypass Feed and Back to Normal; 03/12/2017 - WO 01751083-01; 4KV Relay Routine Calibration (Bus 141 Cub 8) 1A AF Pump; 01/17/2015 - WO 01860298; 1A Diesel Generator Safe Shutdown Sequencer and Single Load Rejection Test; 03/16/2017 - WO 01860299; 1A Diesel Generator SI Signal Override Test 18 Month; 03/14/2017 - WO 01579584; 1B Diesel Generator Safe Shutdown Sequencer and Single Load Rejection Test; 09/28/2015 - WO 01580239; 1B Diesel Generator Sequencer Test; 09/28/2015
| | Changes Feedwater Pump Crosstie Line between Units 1&2 |
| - WO 01582105; 1B Diesel Generator SI Signal Override Test 18 Month; 09/27/2015 - WO 01639571; 2A Diesel Generator Safe Shutdown Sequencer and Single Load Rejection Test; 05/03/2016 - WO 01638485; 2A Diesel Generator Sequencer Test; 05/03/2016 - WO 01638738; 2A Diesel Generator SI Signal Override Test; 05/03/2016 - WO 01635863; 2B Diesel Generator Safe Shutdown Sequencer and Single Load Rejection Test 18 Month; 10/14/2014 - WO 01635582; 2B Diesel Generator Sequencer Test 18 Month; 10/14/2014 - WO 01635773; 2B Diesel Generator SI Signal Override Test 18 Month; 10/14/2014 - Trending Data for Unit 1 and Unit 2, Train A and B, AF006 and AF017 Valves referenced in ER-AA-302-1003, MOV Margin Analysis and Periodic Verification Test Intervals, Revision 8 - Vendor Manual for Auxiliary Feedwater Pump Battery Chargers; Revision 5 - Auxiliary Feedwater System Health Report for Unit 1 and Unit 2 - Chapter 26 11-AF-XL-01; PWR Initial License Training Auxiliary Feedwater System; Revision 11 - Time Critical Action Validation #15; Multiple Spurious AF0005 Valve Actuation; 12/16/2017 - PMID 00190896; Inspect Coupling 1AF01PA-L on O3 frequency - PMID 00103863; Overhaul Pump When Degraded Performance Is noted from Tech Staff - AF-MP-01; Verification of AF System Overpressure Protection; Revision 002 - 53484REL174; Byron 1AF01PA pump bearing oil Analysts Inc report; 10/25/2016 - 53484REL174; Byron 1AF01PA- Outboard Motor Bearing OMB oil Analysts Inc report; 10/25/2016 - 53484REL156; Byron 1AF01PA-Inboard Motor Bearing oil Analysts Inc report; 04/14/2015 - SCI Project 96000075; AMETEK Solidstate Controls Instruction/Operation Manual 10 kVA Rectifier and Inverter 1-E 10 kVA Isolimiter; Revision A - 38092A; Pump Test Data Test Curve (Pacific Pumps); 06/28/1979 - Stock Order 77F14084; Westinghouse Report of Tests on Induction Motors; 07/06/1978 - 953077; Equipment Specification Instrument Bus Distribution Panel; 03/18/1976 - 16-018; UFSAR Change Request Provide Clarification for the Start Time of Passive Leakage Post-LOCA; 09/30/2015 - Byron-01-5048, File 1.10.0101; Memorandum Byron AOV Risk Categorization Document; 08/09/2001 - EC 389382; AF Pumps IST Test Criteria Including Instrument Uncertainty to Support Design Analysis Assumptions; Revision 0 - T340-0067; XOMOX Matryx Vane Actuators; 2009 - 4098736; Troubleshooting Log 1(2)RF026/27; 02/06/2018 - F/L-2795; Specification for Diaphragm Valves (Safety Category I); Amendment 2 - BB-SURV-13; Risk Assessment Missed Surveillance; Revision 0 - IST Program Plan; Inservice Testing Program Fourth Ten Year Interval; 07/01/2016 - Procedural Change Request 1BFR-H.1 Revision 300; Response to Loss of Secondary Heat Sink; 07/30/2017 - 50.59 Screen: VFR-H.1, BFSG-3 Revision 300/300/1/3; Response to Loss of Secondary Heat Sink/Alternate Low Pressure Feedwater Unit; 7/20/2017 - EC 389382; AF Pumps IST Criteria With Instrument Uncertainty for Design Basis Event Flows; 08/07/2012 - EC 389241; Degraded Voltage 5 Minute Timer Resolution (Unit 1); Revision 11 - EC 386525; Design Change to AFW Suction Pressure Logic (Unit 2); Revision 4 - EC 362858; Installation of Unit 1 portion of 1A and 2A MDAFW pumps Unit 1 and 2 cross-tie line; 04/28/2008 - EC 362168; Second Phase final tie-in to Unit 2 for the Unit 1 and 2 MDAFW pumps cross-tie; 10/24/2008
| | Miscellaneous Email AFW Issue - 5054x Licensees_BGH.docx 08/28/2019 |
| - EC 386525; Design Change to AFW Suction Logic; Revision 4 - - EC 378684; Installation of Drain on Shell Side of Jacket Water Cooler (2SX01K) and Installation of Continuous Vent on the Thermostats for 2B AFW Diesel (2AF01PB-K); Revision 2 - 6E-15-118/BRW-S-2015-139; 50.59 Screening - Provide Clarification for the Start Time of Passive Leakage Post-LOCA; Revision 0 Corrective Action Documents Generated as a Result of the Inspection - AR 04096381; DBAI NRC Identified oved Petcocks; 01/23/2018 - AR 04096346; DBAI Housekeeping Issue Identified by NRC; 01/23/2018 - AR 04097122; DBAI NRC Identified Clarification to 1/2 BFR H.1; 01/25/2018 - AR 04099722; 2018 NRC DBAI: Nickel Cadmium Battery Float Voltage Range; 02/01/2018 - AR 04101248; 2018 NRC DBAI: AFW Nickel Cadmium Battery Sizing; 02/06/2018 - AR 04096403; NRC identified oil leak on oil reservoir cover of 1AF01PA; 01/23/2018 - AR 04102086; 2018 DBAI: Fault on NSR Loads Supplied by Instrument Inverter; 02/08/2018 - AR 04096326; NRC Identified Loose Latch on Cabinet Door; 01/23/2018 - AR 04096766; DBAI Identified Check Valve Testing on 1(2)RF026/27; 01/24/2018 - AR 04096407; DBIA Housekeeping 1HS-AF152 Panel Has Loose Clamps; 01/23/2018 - AR 04096409; DBIA Housekeeping Panel For 2HS-AF134 Needs Clamps Tightened; 01/23/2018 - AR 04101772; NRC DBAI EDG Steady State Lower Frequency Admin Limit; 02/07/2018 - AR 04101416; DBAI No Docs to Support IA Regulator Settings for RF026/27; 02/06/2018 - AR 04098736; NRC DBAI - IST Non-Compliance Issue; 01/30/2018 - AR 04097360; DBAI; NRC Identified Housekeeping Issue; 01/23/2018 - AR 04101795; NRC Question on 50.59 Screening for BFR H.1; 02/07/2018
| | Federal Register Applicability of License; Conditions and Technical 04/01/1983 |
| May 7, 2018 Mr. Bryan
| | Vol. 48, No. 64 Specifications in an Emergency |
| : [[contact::C. Hanson Senior VP]], Exelon Generation Company, LLC President and CNO, Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555 SUBJECT: BYRON STATION, UNITS 1 AND 2NRC DESIGN BASES ASSURANCE INSPECTION (TEAMS) INSPECTION REPORT 05000454/2018010; 05000455/2018010 Dear Mr. Hanson: On February 8, 2018, the
| | ML043440415 Letter to A. Scherer (Southern California Edison) from 02/05/1999 |
| : [[contact::U.S. Nuclear Regulatory Commission (NRC) completed a Triennial Baseline Design Bases Assurance Inspection (Teams) at your Byron Station]], Units 1 and 2. On April 12, 2018, the NRC inspectors discussed the results of this inspection with Mr. Kanavos, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report. Based on the results of this inspection, the NRC has identified three issues that were evaluated under the risk significance determination process as having very-low safety significance (Green). The NRC has also determined that three violations are associated with these issues. Because the licensee initiated condition reports to address these issues, these violations are being treated as Non-Cited Violations (NCVs), consistent with Section 2.3.2 of the Enforcement Policy. These NCVs are described in the subject inspection report. If you contest the violations or significance of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the
| | J. Moore (NRC) |
| : [[contact::U.S. Nuclear Regulatory Commission]], ATTN: Document Control Desk, Washington, DC 205550001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement; and the NRC Resident Inspector at the Byron Station. If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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
| | NEI 96-07 Guidelines for 10 CFR 50.59 Implementation 1 |
| : [[contact::U.S. Nuclear Regulatory Commission]], ATTN: Document Control Desk, Washington, DC 205550001; with copies to the Regional Administrator, Region III; and the NRC resident inspector at Byron Station.
| | Procedures 1BFR-H.1 Response to Loss of Secondary Heat Sink 300 |
| : [[contact::B. Hanson -2- 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, Part Inspections, Exemptions, and Requests Sincerely, /RA/ Mark Jeffers, Chief Engineering Branch 2 Division of Reactor Safety Docket Nos. 50454; 50455 License Nos. NPF37; NPF66 Enclosure: IR 05000454/2018010; 05000455/2018010 cc: Distribution via LISTSERV
| | AD-AA-101-1002 Writer's Guide for Procedures and T&RM 18 |
| : [[contact::B. Hanson -3- Letter to Bryan C. Hanson from Mark Jeffers dated May 7]], 2018 SUBJECT: BYRON STATION, UNITS 1 AND 2NRC DESIGN BASES ASSURANCE INSPECTION (TEAMS) INSPECTION REPORT 05000454/2018010; 05000455/2018010
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[Table view] |
Inspection Report - Byron - 2018010 |
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Text
March 19, 2021
SUBJECT:
BYRON STATION - CLOSURE OF AUXILIARY FEEDWATER SYSTEM UNRESOLVED ITEM 05000454/2018010-04 AND 05000455/2018010-04
Dear Mr. Rhoades:
On December 4, 2020, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Byron Station and discussed the results of this inspection with Mr. K. McGuire and other members of your staff. The results of this inspection are documented in the enclosed report.
No findings or violations of more than minor significance were identified during this inspection.
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,
/RA/
Karla K. Stoedter, Chief Engineering Branch 2 Division of Reactor Safety Docket Nos. 05000454 and 05000455 License Nos. NPF-37 and NPF-66
Enclosure:
As stated
Inspection Report
Docket Numbers: 05000454 and 05000455 License Numbers: NPF-37 and NPF-66 Report Numbers: 05000454/2020010 and 05000455/2020010 Enterprise Identifier: I-2020-010-0055 Licensee: Exelon Generation Company, LLC Facility: Byron Station Location: Byron, IL Inspection Dates: April 13, 2020 to December 04, 2020 Inspectors: E. Sanchez Santiago, Senior Reactor Inspector Approved By: Karla K. Stoedter, Chief Engineering Branch 2 Division of Reactor Safety Enclosure
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a design basis assurance inspection (teams) inspection at Byron 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
No findings or violations of more than minor significance were identified.
Additional Tracking Items
Type Issue Number Title Report Section Status URI 05000454, Use of 10 CFR 50.54(x) for 71111.21M Closed 05000455/2018010-04 Unit AFW Cross-Tie
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.
INSPECTION RESULTS
Unresolved Item Use of 10 CFR 50.54(x) for Unit AFW Cross-Tie 71111.21M (Closed) URI 05000454,05000455/2018010-04
Description:
In 2008, the licensee added steps to Emergency Operating Procedure (EOP)1/2BFR-H.1, Response to Loss of Secondary Heat Sink, to use the motor-driven auxiliary feedwater (MDAFW) train of a non-accident unit to combat a loss of all feedwater event in the opposite unit by using a recently installed unit cross-tie. The EOPs also directed operators to enter the technical specification limiting condition for operation action statement for the unit donating the MDAFW train because the MDAFW trains were not licensed to be shared between the reactor units.
In 2011, the resident inspectors documented a Severity Level IV non-cited violation (NCV) of 10 CFR 50.59 after determining the licensees 2008 EOP change resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system or component important to safety. Specifically, the Updated Final Safety Analysis Report described the auxiliary feedwater (AF) trains as non-shared systems. This violation was documented in Inspection Report 05000454/2011004; 05000455/2011004 as NCV 05000454/2011004-02; 05000455/2011004-02, Modification of the Auxiliary Feedwater System Without Prior NRC Approval (REF: ADAMS Accession No. ML113070678).
As a corrective action to this NCV, the licensee removed the steps in EOP 1/2BFR-H.1, "Response to Loss of Heat Sink," that directed the unit cross-tie to be used and removed credit for the cross-tie in the stations Probabilistic Risk Assessment model. On August 8, 2017, the licensee added direction in EOP 1/2BFR-H.1 to use the AF cross-tie by invoking 10 CFR 50.54(x). The procedure change included a note which stated, If at any time it has been determined that restoration of feed flow to any SG [steam generator] is untimely or may be ineffective in heat sink restoration, then the AF crosstie should be implemented per Step 5 (Page 8). The licensee also added a caution which stated, The AF crosstie should be implemented per Step 5 if other attempts to restore feed flow to the SG(s) will not prevent the initiation of feed and bleed. The DBAI team was concerned the note and caution provided direction to initiate the AF unit cross-tie before bleed and feed, rather than the instruction being provided in the actual procedure step.
During the 2018 design basis assurance inspection, the inspectors initiated an unresolved item to document their concerns associated with the licensees actions in response to the 2011 SLIV NCV. Specifically, the inspectors challenged the use of 10 CFR 50.54(x) to implement this change. In addition, the inspectors noted that the licensees 10 CFR 50.59 screening for the procedure change did not include or evaluate the aforementioned note and caution statements to determine whether the 2017 EOP change required prior NRC approval via a license amendment.
Since 2018, the NRC has evaluated the details associated with the above concerns, engaged NRC staff experts, and reviewed licensee procedures and owners group documents related to the technical aspects of the issue to determine whether a performance deficiency or violation occurred. Based on this review, the inspectors did not identify a performance deficiency or violation of requirements.
Specifically, regarding the concern the licensee used 10 CFR 50.54(x) to implement a permanent change, the inspector concluded the 2017 EOP revision which added the note, caution and the step for implementing the AF cross-tie by invoking 10 CFR 50.54(x) did not constitute a change as defined in 10 CFR 50.59, Changes, tests and experiments. Title 10 CFR 50.59 defines change as, a modification or addition to, or removal from, the facility or procedure that affects a design function, method of performing or controlling the function, or an evaluation that demonstrated that intended functions will be accomplished. Also, NEI 96-07, Section 1.2.4, Relationship of 10 CFR 50.59 to 10 CFR 50.2 Design Bases, states, 10 CFR 50.59 controls changes to both 10 CFR 50.2 design basis and supporting design information contained in the Updated Final Safety Analysis Report. The AF cross-tie would be implemented during beyond design basis events and would require invoking 10 CFR 50.54(x), therefore the change falls outside the site's licensing and design basis.
The inspectors also reviewed the URI concern:
- (1) the added note and caution were the only procedure provisions that provided direction on when to use the MDAFW cross-tie; and (2)those procedure provisions were not included in the 10 CFR 50.59 screening. Per the above discussion, the inspectors concluded the change did not fall under the requirements of 10 CFR 50.59. Regarding the caution statement providing direction, the inspectors noted that the caution includes a should statement which per procedure AD-AA-101-1002, Writers Guide for Procedures and T&RM, denotes a management expectation but does not require or direct a specific action. When the licensee was questioned on the purpose of the caution statement, they indicated it was to let the operators know the AF cross-tie should be implemented after all other options are exhausted and feed and bleed was the only remaining alternative. The licensee also stated that per EOP 1/2BFR-H.1, Step 2, "Check if Bleed and Feed is Required," if the criteria described in the procedure for initiating feed and bleed are met, the licensed operators would be required to implement the bleed and feed actions.
There is no allowance to bypass actions related to initiating bleed and feed, once the criteria are met. The licensee also stated the bleed and feed actions would not be delayed to allow implementation of the AF cross-tie. Per the licensee, the added note and caution did not require the use of the cross-tie prior to implementing feed and bleed. Step 5 of procedure 1/2BFR-H.1, "Crosstie Train A AF From Unit 1/2," stated, "Shift Manager has: Determined other heat sink restoration efforts are not available or are untimely; Has implemented 10 CFR 50.54(x); Approved implementation of 1BFSG-3, Alternate Low Pressure Feedwater for AF crosstie." Based on these instructions, the decision to invoke 10 CFR 50.54(x) and implement the cross-tie would be made by the shift manager. The caution also served the purpose of informing the operators that the implementation of the AF cross-tie would require invoking 10 CFR 50.54(x). Per procedure AD-AA-101-1002, one of the purposes of a caution statement is to alert personnel of violations of rules, regulations or work practices. Therefore, the inspectors did not identify a performance deficiency associated with these actions.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
DOCUMENTS REVIEWED
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.21M Engineering 362168 Installation of the Final Phase of the Motor-Driven Auxiliary 0
Changes Feedwater Pump Crosstie Line between Units 1&2
Miscellaneous Email AFW Issue - 5054x Licensees_BGH.docx 08/28/2019
Federal Register Applicability of License; Conditions and Technical 04/01/1983
Vol. 48, No. 64 Specifications in an Emergency
ML043440415 Letter to A. Scherer (Southern California Edison) from 02/05/1999
J. Moore (NRC)
NEI 96-07 Guidelines for 10 CFR 50.59 Implementation 1
Procedures 1BFR-H.1 Response to Loss of Secondary Heat Sink 300
AD-AA-101-1002 Writer's Guide for Procedures and T&RM 18
6
