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05000454/FIN-2018003-0130 September 2018 23:59:59ByronNRC identifiedMinor ViolationOn June 14, 2018, the licensee performed IST surveillance 2BOSR 5.5.8.DO1, Test of the Diesel Oil Transfer System, on the 2A diesel oil transfer pump. On June 19, 2018, the inspectors noted that an issue concerning the calibration of the Flexim ultrasonic flow meter used during the test had not been documented in the licensees Corrective Action Program (CAP). Specifically, the calibration sticker on the flow meter used during the surveillance test indicated that the instrument was calibrated to a 5 percent accuracy when the ASME OM Code required an instrument accuracy of 2 percent. The inspectors discussed the issue with licensee management. The licensee subsequently confirmed that the instrument calibration did not meet ASME OM Code requirements and entered this issue into their CAP.Title 10 CFR Part 50, Appendix B, Criterion XII, Control of Measuring and Test Equipment, requires that measures be established to assure that tools, gages, instruments, and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specific periods to maintain accuracy within necessary limits. Licensee procedure ERAA321, Administrative Requirements for Inservice Testing,Section 4.10.3, states, in part, that instrument accuracy and range requirements are specified in the applicable ASME Code Edition/Addenda. ASME OM Code Paragraph ISTB-3510, General, states, in part, that instrument accuracy shall be within the limits of Table ISTB-35101, Required Instrument Accuracy. Table ISTB35101 states that the required instrument accuracy for determining flow rate is 2 percent. Screening: The failure to implement programmatic controls that ensured measurement and test equipment was calibrated to the accuracy requirements of the ASME OM Code was a performance deficiency. The instruments used in IST surveillances were later re-certified to meet the required 2 percent accuracy in the ASME OM Code with no required adjustments. As a result, the performance deficiency was determined to be minor because the inspectors answered No to all of the more-than-minor screening criteria in IMC 0612, Appendix B. The licensee generated Issue Report (IR) 04149294 to document this issue in their CAP. This issue was also incorporated into a corrective action program evaluation (CAPE) report evaluating an adverse trend identified with ASME test performance at the site (AR 04154533). Violation: The failure to comply with 10 CFR Part 50, Appendix B, Criterion XII, Control of Measuring and Test Equipment, constituted a minor violation that was not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000454/FIN-2018410-0130 June 2018 23:59:59ByronLicensee-identifiedLicensee-Identified Violation
05000454/FIN-2018002-0330 June 2018 23:59:59ByronNRC identifiedMinor ViolationMinor Violation: The inspectors identified multiple instances of a failure to perform inservice testing in accordance with written procedures appropriate for the circumstances during this inspection period: 1. On March 30, 2018, the licensee performed 1BOSR 5.5.8.DO2, Test of the Diesel Oil Transfer System, and declared the 1B diesel oil transfer pump inoperable due to flow results being low out of specification. Subsequently, the licensee determined that the instrument setup was incorrect in that an incorrect value was entered into the flow meter for pipe diameter. The licensee declared the surveillance invalid and scheduled a time to re-perform the activity. Acceptable system flow rates were achieved a week later when the correct pipe diameter was used for the instrument setup. 2. On April 26, 2018, while observing the licensee perform 2BOSR 5.5.8.CS.52C, Comprehensive Inservice Testing (IST) Requirements for Containment Spray Pump 1CS01PB, the inspectors noted that the pump suction pressure and discharge pressure test gauges were not installed as described in the Precautions and Limitations section of the procedure. After the inspectors asked how the installed configuration satisfied the procedure requirement, the licensee suspended the test to obtain clarification. After some deliberation between engineers and operators attempting to identify the correct instrument location, the test data was recorded with the instruments at different locations for data gathering and comparison. The licensee verified that pump performance had sufficient margin, including the introduced error, to remain operable and available to perform its safety-related function as expected.3. On May 1, 2018, while observing the licensee perform 2BOSR 5.5.8.SX.51C, Comprehensive Inservice Testing (IST) Requirements for the Essential Service Water (SX) Pump 2SX01PA and Unit 2 SX Pumps Discharge Check Valves, the inspectors noted that operators were not taking data from the ultrasonic flow meter in accordance with the procedure. Specifically, the instrument was not set up to indicate time and flow so that an average flow could be determined as required by a Note in the procedure. Instead the operators were recording instantaneous flowrate. When the inspector asked for clarification and the operators and technicians deferred to their supervisors, the licensee suspended the test to obtain clarification. The test was performed again after the instrument was set up correctly and operators were briefed on how to obtain the correct data.Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions. Contrary to the above, for the diesel fuel oil transfer pump surveillance, 1BOSR 5.5.8.DO2, the procedure listed an incorrect pipe diameter value that was subsequently entered into the flow meter resulting in unacceptable test results; for the containment spray pump surveillance, 2BOSR 5.5.8.CS.52C, the licensee potentially introduced an unaccounted for error in the surveillance test method by not setting up test equipment in accordance with the procedure; and for the SX surveillance, 2BOSR 5.5.8.SX.51C, the licensee introduced a potential error in the surveillance test by not determining an average flow rate as discussed in the procedure Note.Screening: The failure to perform inservice testing in accordance with written procedures appropriate for the circumstances was a performance deficiencyin each of the listed 11 examples. The performance deficiency was determined to be minor in each case because the inspectors answered No to all of the more-than-minor screening questions in IMC 0612, Appendix B. The licensee generated the following issue reports (IRs) to document these issues:AR 04121539, Ultrasonic Flow Measurement Installation IssueAR 04122295, PCR (procedure change request) 1/2BOSR 5.5.8.DO1 AR 04131201, Engineering Clarification Needed on ASME Precaution AR 04133585, NRC ID: Potential Concerns With Execution of 2A SX Pump Surveillance Violation: These failures to comply with 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, constituted minor violations that are not subject to enforcement action in accordance with the NRCs Enforcement Policy
05000455/FIN-2018002-0130 June 2018 23:59:59ByronSelf-revealingOverspeed Trip of 2B Auxiliary Feedwater Pump During SurveillanceA finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was self-revealed when the 2B diesel-driven auxiliary feedwater (AF) pump tripped on overspeed during a quarterly inservice test (IST). Specifically, operators with portable instrumentation used an erroneous speed value to adjust pump speed beyond the range specified in the procedure resulting in a pump overspeed trip, entry into a 72-hour technical specification (TS) required action statement, and unplanned pump unavailability with an associated change in Unit 2 risk from green to yellow.
05000454/FIN-2018002-0230 June 2018 23:59:59ByronLicensee-identifiedLicensee-Identified Violation

A violation of very low safety significance was identified by the licensee, has been entered into the licensees corrective action program, and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy

Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.Licensee procedure ERAA321, Administrative Requirements for Inservice Testing, stated in Step 4.10.5, that acceptance criteria are established using the reference values and the applicable ASME (American Society of Mechanical Engineers) Code. Paragraph ISTA3160, Test and Examination Procedures, of the ASME Operation and Maintenance of Nuclear Power Plants (OM) Code required in part that, Tests and examinations shall be performed in accordance with written procedures. The procedures shall contain the Owner-specified reference values and acceptance criteria. Paragraph ISTA9230, Inservice Test and Examination Results, of the ASME OM Code required, in part, that The results of tests and examinations shall be documented and shall include the following: comparison with allowable ranges of test and examination values, and analysis deviations and requirements for corrective action.Contrary to the above, from July 1, 2016, to May 30, 2018, the licensees procedures did not clearly document acceptance range, alert range, and required action values for the diesel oil (DO) transfer pump IST surveillance tests in accordance with the ASME OM Code. This resulted in several instances where the pump being tested did not meet IST criteria, but no action was taken. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedural Quality attribute of the Mitigating Systems Cornerstone and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to clearly identify the acceptance criteria, alert range and required action ranges resulted an in organizational failure to declare the pumps inoperable and to perform required analysis of the equipments condition. The inspectors assessed the significance of the finding using SDP Appendix A and concluded the issue was of very low safety significance (i.e., Green).Corrective Action References: (1) AR 04142617, Acceptance Criteria Not Clearly Listed in DO Pump Procedures, and (2) AR 04142370, DO Pump Test Packages are Not Routed to the IST Coordinator.
05000454/FIN-2018010-0131 March 2018 23:59:59ByronNRC identifiedFailure to Prescribe Motor Driven Auxiliary Feedwater Pump Test Procedures that Accounted for the Allowed Emergency Diesel Generator Frequency VariationThe 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, Instructions, Procedures, and Drawings, for the failure to prescribe 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.
05000454/FIN-2018010-0331 March 2018 23:59:59ByronNRC identifiedFailure to Verify the Adequacy of the Air Pressure Regulator SetpointValue for Containment Isolation Valves 1(2)RF026The inspectors identified a Green finding and an associated NCV of 10 CFR Part 50, AppendixB, Criterion III, Design Control, for the failure to verify the adequacy of the air 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 stroketimes.
05000454/FIN-2018010-0231 March 2018 23:59:59ByronNRC identifiedFailure to Periodically Test Instrument Air Check Valves Associated with Air-Operated Containment Isolation ValvesThe inspectors identified a Green finding and an associated NCV of 10 CFR Part 50, AppendixB, Criterion XI, Test Control, for the failure to periodically test 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 closed air-operated containment isolation valves 1(2)RF026 and 1(2)RF027 in response to a containment isolation signal.
05000454/FIN-2018010-0431 March 2018 23:59:59ByronNRC identifiedUse of 10 CFR 50.54(x) for Unit AFW Cross-TieIn 2008, the licensee added steps to Emergency Operating Procedure (EOP) 1/2BFR-H.1, Response to Loss of Secondary Heat Sink, 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; 05000455/2011004-02, Modification of the Auxiliary Feedwater System Without Prior NRC Approval (REF: 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-tie in the stations Probabilistic Risk Assessment model. However, on August 8, 2017, the licensee added direction in EOP1/2BFR-H.1 to use the Unit Auxillary Feedwater cross-tie by invoking 10 CFR 50.54(x). Specifically, the change added a note and a caution that provided direction to initiate the MDAFW unit cross-tie before bleed and feed. The note stated:If at any time it has been determined that restoration of feed flow to any SG is untimely or may be ineffective in heat sink restoration, then the AF crosstie should be implemented per Step 5 (Page 8). The caution 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. Step 5 provided instructions on how to perform the cross-tie and did not include instructions on when to initiate it. The caution also stated Use of the AF crosstie requires invoking 50.54(x).During this inspection period, the inspectors challenged the use of 10 CFR 50.54(x) to implement this permanent change. In addition, the inspectors noted that the licensees 10 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 evaluateif a performance deficiency and or violation occurred. This Unresolved Item will remain open pending the outcome of this ongoing review.
05000454/FIN-2017010-0331 December 2017 23:59:59ByronNRC identifiedFailure to Properly Correct Errors in Design Analysis for Main Steam Line Break in Main Steam TunnelThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly address the identification of secondary missiles following a HELB event. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075608 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating 4 Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000455/FIN-2017004-0131 December 2017 23:59:59ByronSelf-revealingFire Barrier Impaired without AuthorizationA finding of very low safety significance and an associated NCV of Technical Specification 5.4.1.c, Procedures, was self-revealed when an Operations department supervisor identified that a fire door separating two rooms containing safety-related equipment was impaired and did not meet the requirements specified in fire protection program procedures. Specifically, on October 5, 2017, a fire door was left unattended and unable to latch due to the presence of tape over the door latch assembly. The supervisor promptly removed the tape to restore the fire doors functionality and documented the as-found condition in IR 04059911, Fire Door 0DSD474 Improperly Impaired Tape Over Latch. This issue was determined to be of more than minor significance because it was associated with the Initiating Events Cornerstone attribute of Protection Against External Factors (Fire) and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding screened as having very low safety significance (Green) using IMC 0609, Appendix F, Fire Protection Significance Determination Process, Question 1.4.3A, since the fire finding category was determined to be Fire Containment, due to the door not being able to latch, and the combustion loading on both sides of the door was determined to result in less than the 1.5 hour threshold. The finding affected the cross-cutting area of Human Performance in the aspect of Avoiding Complacency (H.12) because the individual that impaired the door did not recognize the inherent risk in their actions and use error reduction tools to mitigate that risk.
05000454/FIN-2017010-0231 December 2017 23:59:59ByronNRC identifiedInadequate Blow Out Panel Design ControlThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee originally designed the MSSV blow out panels in a manner that prevented them 3 from functioning properly. The licensee entered this issue into their CAP as A R 4075608 and corrected the design issue in March of 2009 . The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000454/FIN-2017403-0131 December 2017 23:59:59ByronLicensee-identifiedLicensee-Identified Violation
05000454/FIN-2017010-0531 December 2017 23:59:59ByronNRC identifiedInaccurate Analysis of RecordThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to maintain an accurate and up- to-date analysis of record for a postulated HELB in the MSSV rooms and steam tunnels. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075608 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000454/FIN-2017010-0431 December 2017 23:59:59ByronNRC identifiedUntimely Corrective Action for Secondary MissilesThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly address the identification of secondary missiles following a HELB event. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075608 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating 4 Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000454/FIN-2017010-0131 December 2017 23:59:59ByronNRC identifiedFailure to Prevent Secondary Missiles Following a Postulated HELBThe inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regualtions (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that the design basis for the main steam safety valve (MSSV) room maintenance hatches was maintained. Specifically, the high energy line break ( HELB) analysis performed for the MSSV rooms and steam tunnels prior to initial construction concluded that no secondary missiles were generated as a result of a HELB although maintenance hatches in the ceiling of the MSSV rooms were identified to become secondary missiles following a HELB in the MSSV rooms and steam tunnels. As part of their immediate corrective actions, the licensee entered this issue into their corrective action program (CAP) as AR 4075608 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of system s that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC (Structure, System, and Component) , does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross- cutting aspect was assigned to this finding as it was not reflective of current performance.
05000454/FIN-2017405-0230 September 2017 23:59:59ByronNRC identifiedSecurity
05000454/FIN-2017007-0130 September 2017 23:59:59ByronNRC identifiedFai lure to Perform Maintenance in Accordance with Performance Centered Maintenance TemplateThe inspectors identified a finding of very low safety significance and an associated NCV of TS 5.4.1, Procedures, when licensee personnel failed to perform maintenance in accordance with written procedures as required by Regulatory Guide 1.33. Specifically, from February 3, 2014, through August 25, 2017, the licensee failed to develop and execute work instructions of sufficient scope to accomplish the 3 preventive maintenance to replace flexible hoses on the essential service water (SX) makeup pumps and the diesel driven auxiliary feedwater (AFW) pumps and did not have a technical justification for a deviation from the Exelon Corporate Performance Centered Maintenance (PCM) template. The licensee entered this issue into their CAP as Action Request (AR) 03961955, AR 03971962, and AR 04045769 and planned to replace the flexible hoses at the next available opportunity. The inspectors determined that failure to perform maintenance in accordance with written procedures as required by TS 5.4.1, Procedures, and Regulatory Guide 1.33 was a performance deficiency. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences . Specifically, failing to replace flexible hoses on the SX makeup pumps and the Unit 1 and Unit 2 diesel -driven AFW pumps at a pre - established frequency could allow hose degradation to remain unidentified and lead to the unplanned inoperability of these safety-related systems. Since the finding is a deficiency affecting the design or qualification of mitigating systems, structures and components (SSC s) and the SSC s remained operable and functional, the finding screened as having very low safety significance. This finding affected the C ross -Cutting area of Human Performance in the aspect of Work Management because the licensee failed to perform required maintenance in accordance with their associated maintenance strategy as well as the corporate PCM template (H.5) .
05000454/FIN-2017007-0230 September 2017 23:59:59ByronNRC identifiedFailure to Promptly Identify Degraded Reactor Containment Fan Cooler CircuitryThe inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly identify a condition adverse to quality resulting in a safety -related system becoming inoperable. Specifically, from May 5, 2017, to August 4, 2017, the licensee failed to trend available surveillance data in a timely manner and did not identify a degraded condition in the 1A reactor containment fan cooler (RCFC) time delay circuitry prior to the system becoming inoperable. The licensee entered this issue into their CAP as AR 04039037 and AR 04045767, replaced the failed relay, and planned to update the RCFC system monitoring plan to note abnormal changes in time delay relay actuation times and improve coordination between engineering and operations to reduce the time it takes engineering to obtain RCFC surveillance data for trending after surveillances are completed. The inspectors determined that the failure to promptly identify a condition adverse to quality associated with the time delay relay circuitry in the 1A RCFC was a performance deficiency. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify a degraded condition in the time delay circuitry associated with the 1A RCFC resulted i n a missed opportunity for the licensee to evaluate the cause and initiate prompt actions to respond to the degraded condition prior to the failure. The inspectors answered No to questions A.1 through A.4 of IMC 0609, Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions ; therefore, the finding screened as 4 having very low safety significance. This finding affected the Cross -Cutting area of Problem Identification and Resolution in the aspect of Trending because information was available that indicated a degraded condition in the 1A RCFC time delay relay circuitry for three months prior its failure in August , but was not identified and evaluated by the licensee prior to failure (P.4) .
05000454/FIN-2017405-0130 September 2017 23:59:59ByronNRC identifiedSecurity
05000454/FIN-2017009-0130 June 2017 23:59:59ByronNRC identifiedFailure to Perform 10 CFR 50.59 Evaluation for UFSAR ChangeThe inspectors identified a Severity Level IV, Non-Cited Violation of 10 CFR 50.59, Changes, Tests, and Experiments, Section(d)(1) and an associated finding of very low safety significance (Green) for the licensees failure to provide a written evaluation which provided the basis for the determination that a change did not require a license amendment. Specifically, the licensee failed to provide a basis for why a change to the surveillance frequencies of emergency diesel generators described in the Updated Final Safety Analysis Report did not require prior NRC approval.The inspectors determined that the performance deficiency was more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The associated finding screened to Green (very low safety significance) because it did not result in the loss of operability or functionality. The diesel generators passed their most recent surveillances. As a result the violation is categorized as Severity Level IV in accordance with section 6.1.d of the NRC Enforcement Policy. The issue did not have a cross-cutting aspect because it was not reflective of current performance.
05000454/FIN-2017002-0130 June 2017 23:59:59ByronSelf-revealingFailure to Verify Computer Software during a Transformer Replacement ModificationGreen . A finding of very low safety significance was self -revealed on March 28, 2017, when operators rapidly reduced generator load in response to a loss of forced cooling for the newly installed Unit 1 East main power transformer ( 1E MPT ) and an indicated rapid rise in transformer winding hotspot temperature caused by vendor data entry errors in the monitoring system software . The process detailed in CC -AA- 256- 101, Software Quality Assurance Process for Plant Digital Instrumentation and Control Systems and Components, to verify and validate the software/firmware during updates was not implemented after the vendor made changes to the digital software during the modification process. The issue was entered into the licensees corrective action program (CAP) and corrective actions included replacement of the cooling group supply breaker, correction of the software errors, and revision of the alarm response procedure and supporting documentation. The inspectors concluded that the issue was more than minor because it adversely impacted the Design Control attribute of the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during plant operations. Specifically, rapid power changes or load reject could challenge operating safety limits. In this event, the rapid rise in the calculated winding hotspot indications and subsequent operator actions to rapidly reduce load over 300 megawatts electric ( MWe ) was the result of two software errors : (1) an incorrect Current Turns (CT) Ratio and (2) the incorrect configuration of the MPT cooling groups in series within the software. The inspectors utilized Exhibit 1, Initiating Events Screening Questions of IMC 0609, Significance Determination Process, Appendix A, dated June 19, 2012, to conclude that the finding was Green, or of very low safety significance, because the event did not cause a reactor trip and the event did not affect any mitigation equipment. A cross -cutting aspect in the Challenge the Unknown element of the Human Performance Are a (IMC 0310 H.11) was assigned because the engineering group based the risk evaluation on the vendor input that the scope of the change was limited. The flawed assumption that the vendor input was correct without verification resulted in a failure to manage the risk prior to implementation through the verification/validation of the software/firmware.
05000454/FIN-2017001-0131 March 2017 23:59:59ByronLicensee-identifiedLicensee-Identified Violation

On March 11, 2017 , with Unit 1 shutdown and in a refueling outage, pipefitters as signed to cut out and replace service water valve 1WS413 discovered that piping was blocked upstream of the valve and the work scope was appropriately changed to remove the blocked piping. Taking action they believed was allowed by the work instructions, the pipefitters opened a pipe union and removed the pipe. They then set the removed section containing valve 1WS023C on a nearby tripod to continue work. A system engineer performing a walkdown in the area identified that the removed valve had a clearance (danger) tag on it and immediately stopped work and contacted the operations department. Technical Specification 5.4.1 requires , in part , that written procedures be established, implemented and maintained covering the procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. One administrative procedure recommended in Appendix A is , Equipment Control ( e.g. locking and tagging). OP AA 109 101, Clearance and Tagging, accomplished the locking and tagging requirement for Byron Station. Section 5.2, Danger Tags, established standards for implementation of the tagging process. Step 5.2.2 stated , A component with a Danger Tag attached to it shall not be physically removed from the system. Contrary to the requirements stated above, a component with a danger tag attached was physically removed from the system on March 11, 2017. Specifically, pipefitters disconnected a pipe union and removed associated service water piping from the system that contained valve 1WS023C which had a clearance (danger) tag attached.

The licensee immediately verified that the cooler the piping served was out -of-service on both the supply and return sides with a clearance boundary in place and drained so that the workers were not exposed to a pressurized sourc e. The workers immediately acknowledged their error stating they did not see the tag because they were focused on the demolition activities. The issue was entered into the licensees CAP as IR 03984215 , and the maintenance organization conducted a stand down to reinforce the station standards for compliance with the clearance procedure. The inspectors determined that this issue was more than minor because the performance deficiency adversely impacted the Configuration Control attribute of the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge safety functions during shutdown operations. The inspectors determined the issue was of very low safety significance , or Green by answering No to all screening questions in IMC 0609, Appendix G, Shutdown Operations Significant Determination Process, Exhibit 2, Initiating Events Screening Questions.

05000454/FIN-2016003-0130 September 2016 23:59:59ByronSelf-revealingDOST Flood Barrier Door Left OpenA finding of very low safety significance and an associated NCV of 10 CFR 50 Appendix B, Criterion V, "Instructions, Procedures and Drawings," was self-revealed on September 14, 2016, when a station employee discovered that the flood barrier door for the Unit 1 Train B (1B) diesel oil storage tank (DOST) was open and unattended for three hours and six minutes. The watertight door was installed to protect the DOST fuel oil transfer pumps from the effects of a postulated failure of a circulating water expansion joint at the condenser water boxes in the turbine building, and the open door rendered the 1B diesel generator inoperable. An operator was dispatched to assess the door and, after finding no mechanical issue with the door, closed the door to restore operability to the 1B diesel generator. The issue was entered into the licensees Corrective Action Program (CAP) as IR 02699674. The inspectors determined that the issue was more than minor because it was associated with the Configuration Control attribute of the Mitigating Systems Cornerstone and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to close and dog the 1B DOST door impacted the availability of the 1B diesel generator during postulated events. The finding was determined to be of very low safety significance, or Green, in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Appendix A, The Significance Determination Process (SDP) For Findings at Power, because the inspectors answered the Exhibit 2 Mitigating Systems Screening Question B as Yes. The inspectors determined that the finding involved the degradation of equipment specifically designed to mitigate a flooding event and used Exhibit 4 of the same Appendix to evaluate the significance. The inspectors determined that with the flood door open, this single condition during a turbine building flood event would degrade two trains of a multi-train system. Specifically, the turbine building flood would impact the diesel fuel transfer pumps for both Unit 1 emergency diesel generators. Therefore, a Detailed Risk Evaluation was performed by a Senior Risk Analyst who concluded that the change in core damage frequency (CDF) associated with the finding was 4.6E10/year and since the total estimated CDF was less than 1.0E7/year, the issue screened as having very low safety significance (i.e., Green) using IMC 0609, Appendix H, Containment Integrity Significance Determination Process, for large early release frequency (LERF). The inspectors assigned a cross-cutting aspect in the Avoiding Complacency element of the Human Performance Area (IMC 0310 H.12) to this finding because an individual accessing the room through the doorway failed to challenge the door to ensure proper closure in a manner that would have revealed the door was not properly latched.
05000455/FIN-2016003-0230 September 2016 23:59:59ByronSelf-revealingFailure to Use Alteration Log Resulted in Fuel Oil LeakA finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1.a, Written Procedures, was self-revealed on August 24, 2016, when a fuel oil leak of approximately one-eighth gallon per minute was identified coming from a tubing connection after the Unit 2 Train B (2B) DG was started for routine surveillance testing. Technicians replaced a fuel oil relay during the previous shift and did not use the procedurally required tools to track alterations made to each individual input line as required by MAAA716100, Maintenance Alteration Process. The issue was entered into the licensees CAP as IR 02707888. As part of their corrective actions, the leak was promptly repaired by tightening the fitting after the diesel generator was shut down; and the technicians reviewed human performance error prevention techniques, including proper use of the Maintenance Alterations Log, with supervisors. The inspectors determined that the issue was more than minor because it was associated with the Configuration Control attribute of the Mitigating Systems Cornerstone and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to tighten all fittings during a maintenance activity resulted in a substantial fuel oil leak that could have resulted in a fire or could have impacted the availability of the diesel generator if the tubing had loosened further or become disconnected during a design basis event. The finding was determined to be of very low safety significance, or Green, in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Appendix A, The Significance Determination Process (SDP) For Findings at Power, because the inspectors answered Exhibit 2 Mitigating Systems Screening Question A.1 as Yes since the diesel generator remained operable and functional until the fitting was repaired. The inspectors assigned a cross-cutting aspect in the Avoiding Complacency element of the Human Performance Area (IMC 0310 H.12) to this finding because judicious implementation of human performance error prevention tools could have prevented the failure to properly tighten the fitting, even if the Alterations Log was not used as required.
05000454/FIN-2016201-0130 September 2016 23:59:59ByronNRC identifiedSecurity
05000454/FIN-2016003-0330 September 2016 23:59:59ByronSelf-revealingFailure to Properly Block and Brace a Radioactive Shipment for TransportA finding of very low safety significance and an associated NCV of 10 CFR 71.5(a) and 49 CFR 171.1(b)(12) was self-revealed when the licensee failed to properly block and brace a Radioactive Waste (Radwaste) Shipment that was shipped to a waste processing facility for disposal. The failure to properly block and brace the Radwaste Shipment caused a breach of the shipping package while in transit to the waste processing facility. When the shipment breach was discovered at the waste processing facility, contamination surveys were immediately conducted and it was determined that no loss of content had occurred during transportation. The surveys also determined that radiation dose limits from the package were below NRC and Department of Transportation (DOT) limits. The waste processing facility notified the licensee of the breach during transport and the licensee entered the event into their CAP as IR 02665985. The inspectors determined that the issue was more than minor because it was associated with the Program and Process attribute of the Public Radiation Safety Cornerstone and adversely impacted the cornerstone objective of ensuring adequate protection to public health and safety from exposure to radiation from routine civilian nuclear operations. Specifically, the breach of the transportation package by its content could lead to the inadvertent spread of radioactive contamination to the public domain if conditions had been slightly altered. The finding was determined to be of very low safety significance, or Green, in accordance with IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, dated February 12, 2008, because the finding did not involve: (1) a radioactive shipment above radiation limits; (2) a certificate of compliance issue; (3) the failure to make emergency notifications; or (4) a low-level burial issue. A breach of the transportation package occurred during transit. However, the shipment contained less than a Type A quantity of material (LSA II shipment), and there was no loss of package contents or radioactive contamination. The inspectors assigned a cross-cutting aspect in the Resources element of the Human Performance Area (IMC 0310 H.1) to this finding due to inadequate procedures.
05000455/FIN-2016007-0130 September 2016 23:59:59ByronLicensee-identifiedLicensee-Identified ViolationThe licensee identified a finding of very-low safety significance (Green) and associated Non-Cited Violation of Technical Specification 5.4.1.c which required that written procedures shall be established, implemented, and maintained covering the Fire Protection Program implementation. Procedure MA-BY-EM-1-FP003-002, Diesel Generator and Day Tank Room Low Pressure CO2 System Detection Test, was used by the licensee to test and calibrate the time delay relay that controls the time that carbon dioxide is discharged in the 2A diesel generator room when the carbon dioxide suppression system is actuated. The licensee's Fire Hazards Analysis for the 2A diesel generator room, Fire Zone 9.2-2, stated that the total flooding carbon dioxide system would deliver a sufficient quantity of carbon dioxide to maintain a 34 percent concentration for 10 minutes. Calculation BYR 97-041 established that 70 seconds was required to achieve a 34 percent concentration of carbon dioxide in the diesel generator room, and the preservice test for the carbon dioxide system for that room established that a discharge time of 99 seconds was required to ensure a 34 percent concentration for 10 minutes. Contrary to the above, from January 30, 1987, when the Byron, Unit 2 operating license was issued, until October 19, 2015, when the procedure was revised, the licensee failed to maintain a procedure that verified the capability of the carbon dioxide system. Specifically, Procedure MA-BY-EM-1-FP003-002 directed the maintenance technicians to verify the time delay relay was set between 60 and 80 seconds when the licensee's calculations required 70 seconds to achieve the required carbon dioxide concentration in the room and 99 seconds to maintain that concentration for 10 minutes. The performance deficiency was determined to be more-than-minor because the issue adversely impacted the Mitigating Systems Cornerstone objective to ensure the capability of systems that respond to initiating events and prevent undesirable consequences due to external events such as fire. Specifically, the procedure allowed the carbon dioxide system to be calibrated such that it might not have provided sufficient carbon dioxide to extinguish a fire in the 2A diesel generator room. The inspectors screened the finding using Inspection Manual Chapter 0609, Significance Determination Process, Appendix F, Fire Protection Significance Determination Process. Since the reactor was still able to reach and maintain a SSD condition, the finding screened as very-low safety significance (Green). The licensee entered the issue into the CAP as Issue Report 2571839, 2A Diesel Generator Room CO2 Discharge Time, and revised the procedure to require a 100 second discharge time.
05000454/FIN-2016002-0530 June 2016 23:59:59ByronNRC identifiedLicensee Implementation of Enforcement Guidance Memorandum 15002, Enforcement Discretion for Tornado-Generated Missile Protection NoncomplianceOn June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection, focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliance that had been identified through different mechanisms and referenced enforcement guidance memorandum (EGM) 15002 which was also issued on June 10, 2015. The EGM provided guidance to allow the NRC staff to exercise enforcement discretion when an operating power plant licensee did not comply with the current licensing basis for tornado-generated missile protection. Specifically, the discretion applied to SSCs declared inoperable resulting in TS LCOs that would require a reactor shutdown or mode change if the licensee could not meet the required actions within the TS completion time. The discretion allowed the licensee to re-establish operability through compensatory measures and established criteria for continued operation of the facility as longer term corrective actions were implemented. The EGM stated that the bounding risk analysis performed for this issue concluded that this issue was of low risk significance and, in Byrons case, provided for enforcement discretion of up to three years from the date of issuance of the EGM. However, the EGM did not provide licensees with enforcement discretion for any related underlying technical violations; and moreover, the EGM specifically requires that any associated underlying technical violation(s) be assessed through the enforcement process. Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants (GDC), Criterion 4, Environmental and Dynamic Effects Design Basis, stated in part that SSCs important to safety shall be adequately protected against dynamic effects including missiles. On May 25, 2016, the licensee initiated IR 02673848, identifying a nonconforming condition of Criterion 4. Specifically, multiple locations were identified in the refueling water storage tank (RWST) roof hatches and in the L-line wall above the 451 elevation (separating the turbine building from the Class I auxiliary building) where SSCs were not adequately protected from tornado-generated missiles. The licensee declared multiple SSCs inoperable and promptly implemented compensatory measures designed to reduce the likelihood of tornado-generated missile effects. The inspectors reviewed the licensees compensatory measures that included: review and revision of procedures for a tornado watch and a tornado warning to provide additional instructions for operators preparing for tornados and/or high winds, and a potential loss of SSCs vulnerable to the tornado missiles; confirmation of readiness of equipment and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX); verification that training was up to date for individuals responsible for implementing preparation and response procedures; and establishment of a heightened station awareness and preparedness relative to identified tornado missile vulnerabilities. The condition was reported to the NRC as Event Notice (EN) 51958 as an unanalyzed condition and potential loss of safety function. The licensee documented the inoperability of the SSCs and the affected TS LCO conditions in the CAP and in the control room operating log. The shift manager notified the NRC resident inspector of implementation of EGM 15002, and documented the implementation of the compensatory measures to establish the SSCs operable but nonconforming prior to expiration of the LCO required action. The enforcement discretion was applied to the required shutdown actions of the following TS LCOs for both units: TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs) TS 3.3.7, Control Room Ventilation (VC) Filtration System Actuation Instrumentation; TS 3.5.2, ECCS Operating; TS 3.5.4, Refueling Water Storage Tank (RWST); TS 3.6.6, Containment Spray and Cooling Systems; TS 3.7.9; Ultimate Heat Sink; TS 3.7.10, Control Room Ventilation (VC) Filtration System; TS 3.7.11, Control Room Ventilation (VC) Temperature Control System; TS 3.8.4, DC Sources Operating; TS 3.8.7, Inverters Operating; and TS 3.8.9, Distribution Systems Operating. The inspectors review addressed the material issues in the plant, and whether the measures were implemented in accordance with the guidance documentation for the EGM. The inspectors also evaluated whether the measures as implemented would function as intended and were properly controlled. The licensee implemented actions to track the more comprehensive actions to resolve the nonconforming conditions within the required 60 days. These comprehensive actions were to remain in place until permanent repairs were completed, which for Byron were required to be completed in three years, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC such that discretion was no longer needed. The inspectors did not review the underlying circumstances that resulted in the TS violations. As stated in the EGM guidance, violations of other requirements, including 10 CFR 50 Appendix A, Criterion 4, which may have contributed to the TS violations, would be evaluated independently of the EGM implementation. This operability inspection constituted a partial sample as defined in IP 71111.1505 since all corrective actions to support continued operability and resolution of the nonconforming conditions had not been identified. These actions and any underlying technical violations will be addressed with the completion of this inspection sample and documented in a future NRC Inspection Report.
05000454/FIN-2016404-0130 June 2016 23:59:59ByronLicensee-identifiedLicensee-Identified Violation
05000454/FIN-2016002-0130 June 2016 23:59:59ByronNRC identifiedFailure to Perform ASME Code Case Required Extent of Condition to Identify Unacceptable Piping FlawsA finding of very low safety significance was identified by the inspectors when, upon identification of a through-wall leak, the licensee declared the structural integrity of Class 3 fire protection piping to be operable, but failed to perform augmented examinations within 30 days as required by American Society of Mechanical Engineers (ASME) Code Case N5133. The licensee repaired the leaking pipe, and upon identification by the inspectors, documented the issue in their corrective action program (CAP) as IRs 2639930 and 2652145, and performed the required augmented examinations. The inspectors determined the performance deficiency was more than minor because, if left uncorrected, the finding had the potential to lead to a more significant safety concern. Specifically, the augmented examinations identified a location where wall thickness measurements were below the acceptance criteria such that the pipe could have ruptured during a seismic event, impacting the functionality of the fire protection system and causing a flooding hazard in the auxiliary building. Because the finding involved an internal flooding hazard, a detailed risk evaluation was performed, which determined the finding to be of very low safety significance. The inspectors determined the finding had a cross-cutting aspect in the Problem Identification and Resolution area of Evaluation (P.2), because the licensee failed to thoroughly evaluate the issue to ensure that the resolution addressed the cause and extent of condition commensurate with the safety significance. Specifically, the licensee failed to complete the N-513-3 evaluation and perform the required extent of condition activities in a timely manner as specified by the ASME Code Case.
05000454/FIN-2016002-0430 June 2016 23:59:59ByronLicensee-identifiedLicensee-Identified ViolationTechnical Specification 5.4.1.c requires that written procedures shall be established, implemented, and maintained covering the Fire Protection Program implementation. Step 4.2.9 of OPMW201007, Fire Protection System Impairment and Control, stated, Compensatory measures for inoperable fire protection SSCs shall be established in accordance with site specific TRM (or equivalent document) and impairment procedures. TRM LCO 3.10.g requires implementation of an hourly fire watch or establishment of alternate compensatory measures. Contrary to the above, an hourly fire watch was not established for the lower cable spreading room on March 11, 2016, when the fire suppression system was removed from service and the requirements of the alternate compensatory measure were no longer satisfied. Specifically, the alternate compensatory measure required the suppression and detection systems to be available and when the requirements were no longer satisfied, the hourly fire watch should have been re-established. The condition existed for approximately 5 hours and fire detection remained operable during the entire period. The licensee entered this issue into their CAP as IR 2639686. The performance deficiency was determined to be more than minor because it adversely impacted the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences due to external events such as fire. Specifically, the failure to implement the analyzed compensatory measures reduced the reliability of the systems required for safe shutdown. The inspectors screened the finding using IMC 0609, Significance Determination Process, Attachment 04, Initial Characterization of Findings, Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions. The Part B and D.1 questions were answered No, but question D.2 was answered Yes. Since the detection system remained operable during the entire period, one of the D.2.a conditions was satisfied and the condition represented a finding of very low safety significance (Green). The issue was also reviewed using IMC 0609 Appendix F resulting in a delta CDF of 4.4E7/year which also screened as having very low safety significance (Green).
05000454/FIN-2016002-0330 June 2016 23:59:59ByronLicensee-identifiedLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, 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 calculation methods, or by the performance of a suitable testing program. Contrary to the above, the licensee identified that they failed since original plant construction to verify the adequacy of the diesel driven AFW pump design. Specifically as discussed in the review of LER 05000454/201600100 in Section 4OA3 of this report, the licensee failed to verify the diesel driven AFW pump could perform its safe shutdown function following a HELB in the turbine building. Since the diesels air intake was located in the Turbine Building, it would be impacted by a HELB. The licensee entered this issue into their CAP and took immediate corrective actions by declaring both the Unit 1 and Unit 2 diesel driven AFW pumps inoperable and then restored operability of the pumps by implementing temporary plant modifications to relocate the diesel air intakes to the auxiliary building where the environment was not susceptible to a HELB. The licensees planned corrective actions include a permanent plant modification to relocate the air intake to a location that was not susceptible to a HELB. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to verify that the diesel driven AFW pump could perform its safe function following a HELB event in the turbine building did not ensure its availability, reliability, and capability to respond to the initiating event. Since the finding did represent an actual loss of function of at least a single Train for greater than its Technical Specification Allowed Outage Time, a Detailed Risk Evaluation was performed which concluded that the estimated change in core damage frequency was approximately 3.4E7/year, which represented a finding of very low safety significance (Green).
05000454/FIN-2016405-0130 June 2016 23:59:59ByronNRC identifiedSecurity
05000454/FIN-2016002-0230 June 2016 23:59:59ByronSelf-revealingFailure to Comply With Radiation Work Permit Requirements Resulting In An Unplanned Dose Rate AlarmA finding of very low safety significance and an associated Non-Cited Violation (NCV) of Technical Specification 5.4.1 was self-revealed when an engineer violated a radiation work permit by entering an area that was outside of the scope of the radiation work permit (RWP), which resulted in the engineer receiving an unplanned electronic dosimeter dose rate alarm. After the engineer received the unplanned dose rate alarm, he immediately exited the area and reported the event to the radiation protection staff. The licensee entered this issue into their CAP as IR 02655195. The inspectors determined that the performance deficiency was more than minor because the finding impacted the Program and Process attribute of the Occupational Radiation Safety Cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, the engineer, by entering an area that he was not briefed to enter on the radiation work permit, removed a barrier that was intended to prevent workers from receiving unexpected dose. The finding was determined to be of very low safety significance in accordance with Inspection Manual Chapter (IMC) 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008. The violation was determined to be of very low safety significance (Green) because: (1) it did not involve as-low-as-reasonably-achievable (ALARA) planning or work controls; (2) there was no overexposure; (3) there was no substantial potential for an overexposure; and (4) the ability to assess dose was not compromised. The inspectors determined that the finding had a cross-cutting aspect in the Human Performance area of Challenging the Unknown (H.11) because the individual did not stop when faced with an uncertain condition. Specifically, risks were not evaluated and managed before proceeding.
05000454/FIN-2016001-0231 March 2016 23:59:59ByronSelf-revealingEntry into Mode 3 with Turbine Trip Function of SSPS DisabledA finding of very low safety significance and associated non-cited violation (NCV) of Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.4 was self-revealed when the licensee transitioned Unit 1 to Mode 3 with the turbine trip function of the Solid State Protection System (SSPS) disabled although the turbine trip function was required by TS LCO 3.3.2 to be operable in Mode 3. Upon identification, the licensee immediately manually tripped the turbine and restored the automatic turbine trip function. The licensee entered the issue into the corrective action program (CAP) and initiated actions to revise the mode change checklist and affected surveillance procedures. The inspectors determined that the finding was of more than minor safety significance because it was associated with the Configuration Control aspect of the Initiating Events Cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions. The finding was Green because the manual turbine trip and main steam isolation functions were not affected by the finding. The inspectors determined that the finding had a cross-cutting aspect of Work Management in the area of Human Performance (H.5) because the licensee failed to plan, control, and execute work activities such that nuclear safety was the overriding priority.
05000454/FIN-2016001-0131 March 2016 23:59:59ByronNRC identifiedFailure to Enter Technical Specification Limiting Condition for Operation Action Requirement with Auxiliary Feedwater Flow Control Valves Failed OpenThe inspectors identified an Unresolved Item (URI) associated with the concern that the licensee failed to enter a TS LCO action requirement when all air was isolated to the actuators to the auxiliary feedwater flow control valves, failing them open and unable to be throttled or closed from the control room. Description: On January 3, 2016, the licensee generated Issue Report (IR) 2607148 which requested clarifying guidance from engineering for assessing operability of the 1/2AF005AH auxiliary feedwater flow control valves to the steam generators, when air is isolated from the valve actuators. The IR stated that when procedure BISR 3.4.2200, Surveillance Calibration of Aux Feedwater to Steam Generators A, B, C and D Flow Control Loops, was performed, all air was isolated from the auxiliary feedwater flow control valves to fail them open during the calibration. This was intended to maintain operability of the auxiliary feedwater system during the calibration. At each Byron unit, each of the two auxiliary feedwater pumps had a separate flow path to each of the steam generators, and each flow path had an air-operated flow control valve, a motor-operated containment isolation valve, and a check valve in the flow path. The flow control valves used instrument air as the motive force to throttle and close the valves. Upon a loss of air to the actuator, the flow control valves were designed to fully open via spring pressure, allowing auxiliary feedwater flow to the steam generators. In 2012, the licensee installed safety-related accumulators on each auxiliary feedwater train to supply air to the auxiliary feedwater flow control valve actuators upon a loss of instrument air. This air supply was designed so that if one of the steam generators experienced a steam generator tube rupture and the containment isolation valve in the flow path to that steam generator failed to close, the control room operators could close the flow control valve to limit or isolate auxiliary feedwater flow to the failed steam generator until an equipment operator could locally secure the flow control valve in its closed position. This modification was performed to support the licensees license amendment request for a measurement uncertainty recapture uprate so that operator actions could be credited to prevent the steam generator with a ruptured tube from overfilling and challenging the containment function. Upon completion of the modifications, the licensee updated Table 15.07, Plant Systems and Equipment Credited for Transients and Accident Conditions, in the Accident Analysis section of the licensees UFSAR to include the AF Accumulator Tanks as engineered safeguard feature (ESF) equipment credited for steam generator tube rupture incidents. The safety evaluation in Chapter 10 of the UFSAR was also updated for the Auxiliary Feedwater System to state that in the event of a steam generator tube rupture, operator action was required to isolate auxiliary feedwater flow to the ruptured steam generator within certain time requirements, and that in the event that the containment isolation valve failed to close, the flow path could still be isolated by closing the AF005 valves, with air accumulators sized to ensure sufficient time for local operator action to secure the AF005 valves in the closed position. In response to IR 2607148, the licensees regulatory assurance department documented that the safety function of TS LCO 3.7.5, Auxiliary Feedwater System, was intended to be limited to supply water to the steam generators for heat removal and that this should not be changed in favor of any UFSAR design analysis. The IR concluded that Operability per the TS was not applicable, and operations did not need to place the unit in any TS condition statement with the AF005 valves failed open with no instrument air supply during the associated instrument calibrations. From the time the licensee received the measurement uncertainty recapture license amendment on February 7, 2014, through March 22, 2016, the licensee had failed open all four AF005 valves in each train of auxiliary feedwater using BISR 3.4.2200 at least five times per train, and has not entered a TS LCO during most of these time periods. The regulations stated in 10 CFR 50.26(c)(2)(ii)(C) that a TS LCO of a nuclear reactor must be established for each SSC that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that assumes the failure of/or presents a challenge to the integrity of a fission product barrier. The inspectors were concerned that when all air is isolated to the auxiliary feedwater flow control valve actuators, operators may not be able to throttle or isolate flow to a ruptured steam generator quickly enough to prevent overfill of the steam generator, assuming the motor-operated containment isolation valve fails to close, which could challenge the integrity of containment. As such, the inspectors were concerned that the licensee failed to enter a TS LCO action requirement when the air to the actuator was isolated. To determine whether a performance deficiency or violation exists, the inspectors need to determine if a TS LCO should have been established for the ability of the AF005 valves to close to mitigate a steam generator tube rupture event, and if the licensees modifications and license amendment requests properly addressed the establishment of an LCO for this function of the SSC. (URI 05000454/201600101, 05000455/201600101; Failure to Enter Technical Specification Limiting Condition for Operation Action Requirement with Auxiliary Feedwater Flow Control Valves Failed Open)
05000454/FIN-2016406-0131 March 2016 23:59:59ByronLicensee-identifiedLicensee-Identified Violation
05000454/FIN-2015004-0131 December 2015 23:59:59ByronNRC identifiedInaccurate Technical Basis for Operability Evaluation of Reactor Head Flange DamageThe NRC inspectors identified a finding of very low safety significance (Green) when licensee personnel failed to ensure accuracy of calculations used to support an operability evaluation of the Unit 1 reactor vessel head flange for the impression caused by an allen wrench trapped between the stud tensioner and the head flange during stud de-tensioning. The licensee entered this issue in its CAP as Issue Report 02559542. Corrective actions included a significant revision to the Operability Evaluation to address each of the inspectors concerns. The finding was determined to be more than minor because it was associated with the Reactor Coolant System (RCS) Equipment and Barrier Performance attribute of the Barrier Integrity Cornerstone, and adversely affected the cornerstone objective of providing reasonable assurance that physical barriers RCS protect the public from radionuclide releases caused by accidents or events. Additionally, More than Minor Example 3.a of IMC 0612, Appendix E, Examples of Minor Issues, was used to answer this more than minor screening question. Specifically, the licensee used incorrect area in the bearing stress calculation that, at the time of discovery, resulted in reasonable doubt of the operability as the bearing stress exceeded the allowable stress value used in the evaluation to preclude plastic deformation. In accordance with IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, Table 2, the RCS boundary issues need to be considered under the Initiating Event Cornerstone. Using Table 3, the inspectors determined the finding pertained to an event or degraded condition while the plant was in shutdown and, therefore, used IMC 0609, Appendix G Shutdown Operations Significance Determination Process, dated May 9, 2014, for significance determination. The finding did not represent a loss of level control per the Criteria in Appendix G, Attachment 1. The inspectors reviewed Appendix G, Attachment 1, Exhibit 2, Initiating Events Screening Questions. The inspectors answered No to Question A.1, and found all other questions to be not applicable and, therefore, concluded that the finding was of very low safety significance (Green). This finding had a cross-cutting aspect in Human Performance Avoid Complacency because the licensee reviewer, expecting acceptable results, did not use appropriate rigor in evaluating possible errors. Specifically, the licensee did not expect a numerical error in the evaluation performed by the vendor and did not take expected actions to verify accuracy. (H.12)
05000454/FIN-2015004-0631 December 2015 23:59:59ByronLicensee-identifiedLicensee-Identified ViolationAs discussed in Section 4OA3.1 of this report, the licensee identified through communication of operating experience from Braidwood Station that design deficiencies in the circuits associated with the Pressurizer PORV block valves might have resulted in the valves not being available when required due to fireinduced failures in the event of a fire in the control room, cable spreading rooms, or electrical cable penetration areas. Byron License Conditions 2.C(6) for Unit 1 and 2.E for Unit 2, required, in part, that the licensee implement and maintain in effect all provisions of the fire protection program as described in the FSAR, as supplemented and amended, and as approved in SERs and their supplements. Section 2.4.3.2, Pressurizer PORVs and Block Valves of the Safe Shutdown Analysis, stated, in part, that the Division 12(22) PORV and block valves both have control cables in the main control room and in two of the lower cable spreading rooms. Should a fire in any of these zones cause the spurious opening of the PORV, coincident with control circuit damage to the block valve, the block valve could still be closed. A remote/local isolation switch and control switch are provided for the block valve at its motor control center, located in the Division 12(22) electrical penetration area. The block valve can be closed by placing the remote/local isolation switch in local and then closing the valve with the control switch provided. Additionally, Section 2.4.3.2 also stated that in fire zones where one of the PORVs had a control cable present in the zone that can spuriously open the PORV and its associated block valve does not have AC power available, the PORV will be failed closed by pulling its control power fuse at its DC distribution panel. Contrary to the above, as of August 20, 2015, the licensee failed to implement and maintain all provisions of their approved fire protection program. Specifically, the licensee failed to ensure that control circuits associated with the PORVs and local control function would close the PORV block valve during the postulated fire. The licensee entered this issue into their CAP, established fire watches, and performed plant modifications to correct the issue. The inspectors determined that the issue was more than minor because the performance deficiency impacted the Protection Against External Factors Attribute of the Mitigating Systems Cornerstone in that fire-induced circuit failures could impair the operation of the PORV block valves and complicate shutdown of the plant in the event of a fire in the control room, cable spreading rooms, or electrical cable penetration areas. The finding was determined to be of very low safety significance (Green) based on a detailed risk-evaluation performed by a Region III Senior Reactor Analyst.
05000454/FIN-2015004-0731 December 2015 23:59:59ByronLicensee-identifiedLicensee-Identified ViolationThe licensee identified an NCV of TS 5.4.1, Procedures, for a failure to wear dosimetry as prescribed on the radiation work permit (RWP). The licensees TS 5.4.1 required, in part, that written procedures shall be established, written, and maintained for Access Control to Radiation Areas including a RWP System. Station, procedure RP-AA-1008, Revision 4, Unescorted Access to and Conduct in Radiologically Controlled Areas, requires that radiation workers are responsible to read, understand, and acknowledge the appropriate copy of the RWP for any work requiring an RWP. The RWP 10017249, Revision 0, states that Workers shall be evaluated for proper dosimetry placement. Only Radiation Protection shall reposition dosimetry. Additionally, RP-AA-1008, Revision 4, Unescorted Access to and Conduct in Radiologically Controlled Areas, Step 4.2.7, states, WEAR the primary (DLR) and secondary (electronic) whole-body dosimeters within 6 inches or less (about a hands width) of each other on the chest region unless otherwise specified by Radiation Protection Supervision or the RWP. Dosimetry movement is not allowed unless directed by Radiation Protection. Dosimetry should be facing outward. Contrary to the above, on September 23, 2015, a contract carpenter removed his electronic alarming dosimeter and the dosimetry movement was not directed by Radiation Protection. The licensee entered this issue into the CAP as IR 02559980. The inspectors determined that this issue was of very low safety significance (Green) after reviewing IMC 0609, Appendix C, Occupational Radiation Safety SDP, dated August 19, 2008. The inspectors determined that it was not an ALARA planning issue, there was neither overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. Therefore, the finding screened as Green (very-low safety significance).
05000454/FIN-2015004-0231 December 2015 23:59:59ByronSelf-revealingMispositioned Valve in Diesel Fuel Oil Transfer Pump Recirculation Flow PathA finding of very low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed on October 7, 2015, when the Unit 1 diesel oil storage tank (DOST) high level alarm and 1B DOST sump high-high alarms actuated as a result of a mispositioned valve in the diesel fuel oil (DO) system. Specifically, when administrative controls were removed from two valves in the DO system, one of the valves was not placed in its standby position resulting in fuel oil trains being cross-tied across divisions. The licensee entered this issue into its CAP. Corrective actions included closing the mispositioned valve and restoring fuel oil storage tank levels in both trains. The operators were briefed on the requirement to use controlled documents and using human performance error reduction techniques when identifying the restoration position of components under administrative controls. The inspectors evaluated the performance deficiency in accordance with IMC 0612, Appendix B, Issue Screening, and characterized the issue as more than minor because the performance deficiency is associated with the Mitigating Systems Cornerstone objective attribute of Configuration Control of operating equipment, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to an initiating event. Specifically, mispositioning the 1DO055A so that the fuel oil trains were cross-tied created a flow path during operation of the 1A DG that transferred fuel oil out of the A train tanks to the B train tanks. In this instance, tank low level alarms were received and the senior reactor operators declared the 1A DG inoperable, but operators were able to terminate the event before the tank level reached actual TS minimum level. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 4, Initial Screening and Characterization of Findings, dated June 19, 2012, and IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, Exhibit 2 Mitigating Systems Screening Questions Section A. All questions were answered No. Therefore, the finding screened as Green. The inspectors determined that this finding had an associated cross-cutting aspect in the area of Human Performance Design Margins in that the supervisor assumed the open position was changed by the modification and did not use the appropriate rigor to identify the required position using controlled documents and thereby implementing the design requirements to maintain margin (H.6).
05000454/FIN-2015004-0431 December 2015 23:59:59ByronLicensee-identifiedLicensee-Identified ViolationOn September 21, 2015, during the Unit 1 refueling outage, welders performed a modification to install a new valve on the combined discharge of the B train fuel oil transfer pumps. During the post maintenance run of the 1B diesel generator and support systems (including fuel oil), operators observed high discharge pressure on the fuel oil transfer pumps. Troubleshooting revealed the purge dam material used during the weld application had been wadded up when installed and left in the piping after the welding was complete, had wedged in the 1B diesel fuel oil day tank inlet isolation valve, 1DO005B, and blocked fuel oil flow to the day tank. Welders had assumed that the water-soluble purge dam material (referred to a rice paper) would dissolve in the fuel oil and, therefore, did not need to be removed prior to clearance order release. Several steps in CC-AA-501-1026, Exelon Nuclear Welding Program Purging Techniques, allow use of water-soluble purge dams specified by brand name in recommended applications, but step 4.5.1 specifies that if water-soluble purge dams are used, that they be flushed from the system, and step 4.5.3 assigns responsibility to the supervisor to assure, in part, that all purge materials have been removed from the system. In addition, step 4.2.5.2 states Do not wad water soluble paper purge dam materials into a pipe or tube. Work order instructions did not incorporate the information provided in CC-AA-501-1026 and the weld data sheet only specified that a dam was to be used. Failure to provide work instructions appropriate to the circumstances is a performance deficiency. Title 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be documented by instruction, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with those instructions, procedures, or drawings. Contrary to the above, the work instructions included in WO 01635687 on September 21, 2015, did not include instructions appropriate to the circumstances in that the work order did not include steps to ensure that the purge dam material was removed after welding was complete. The issue was entered into CAP as IR 02559056 and the purge dam material was removed. An extent of condition review determined that other welding performed during this outage and the previous Unit 2 outage was performed in compliance with the guidance included I CC -AA-501-1026. The inspectors determined this issue was more than minor because the performance deficiency impacted the Equipment Performance attribute of the Mitigating Systems Cornerstone in that the fuel oil system was made available for service and the diesel generator was credited as available by the operating staff when, in fact, it was not available. The inspectors used IMC 0609, SDP, Appendix G, Shutdown Operations, Attachment 1, Phase 1 Operational Checklist for Both PWRs and BWRs, Checklist 2, PWR Cold Shutdown Operation, to determine that no quantitative assessment was required and that the issue was Green or very low safety significance.
05000454/FIN-2015004-0331 December 2015 23:59:59ByronSelf-revealingFailure to Implement Protective Tagging Procedure RequirementsA finding of very low safety significance and associated NCV of Technical Specification (TS) 5.4.1.a, Procedures, was self-revealed during the Unit 1 refueling outage that ended on October 2, 2015, as a result of the licensees failure to implement the requirements of OP-AA-109-101, Clearance and Tagging Program. Two instances of personnel failing to implement the procedural requirements were identified. First, on September 18, 2015, workers in the switchyard performed a preventative maintenance task to replace the breaker and removed the old breaker with the danger tag still attached. Additionally, on September 28, a deficient clearance order for the Unit 1 polar crane was put in place to support maintenance, and the clearance order did not incorporate temporary plant configuration changes. The licensee entered both issues in the Corrective Action Program (CAP). The site performed a work stand down with switchyard workers to reinforce the procedural requirements following the first issue and with all operators qualified to prepare and approve clearance orders to communicate the second event, potential consequences, and procedural implementation shortfalls. The site also performed a review of all open temporary configuration changes with clearances to ensure equipment was properly tagged out. The inspectors determined that the licensees failure to implement the requirements of OP-AA-109-101, Clearance and Tagging Program, was a performance deficiency. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined that the issue was more than minor because, if left uncorrected, the performance deficiency could result in a more significant safety concern. Specifically, failure to implement the requirements of the protective tagging program could result in a direct challenge to nuclear safety through an initiating event, barrier degradation or damage to equipment necessary to mitigate an event. The inspectors determined that while the Initiating Events Cornerstone attributes of Equipment Performance and Human Error best addressed the specific performance deficiencies identified, more than one cornerstone was potentially affected since the performance deficiency affected programmatic control of equipment configuration. The inspectors utilized IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, dated May 9, 2014, to evaluate the significance. After evaluating plant conditions at the time the examples occurred, the inspectors used Attachment 1, Phase 1 Initial Screening and Characterization of Findings, Exhibit 2, Initiating Events Screening Questions, and answered all of the questions such that the issue was screened as Green or very low safety significance. The common element to these two examples was the lack of familiarity of the individuals with the process and their understanding of the indications present. As a result, inspectors assigned a Human Performance cross-cutting aspect of Training (H.9).
05000455/FIN-2015004-0531 December 2015 23:59:59ByronLicensee-identifiedLicensee-Identified ViolationOn October 23, 2015, during a return to full power after power maneuvering on October 23, 2015, the Unit 2 axial offset (AO) exceeded the procedural limit of the reactivity maneuver guidance sheet (ReMA). AO is an indication used to ensure power distribution and fuel conditioning limits are properly maintained throughout the core during steady state conditions and when maneuvering the plant up or down in power. The Unit 2 reactor operator (NSO) was focused primarily on RCS temperature control, which had stabilized because the reactivity changes from power ascension and poison burnout were offset enough to stabilize temperature control, and did not recognize the AO trend was still becoming more negative. The operator was not monitoring all of the critical parameters specified in the ReMA and, as a result, the AO value dropped to - 4.3% and exceeding the 3% from a target value of -0.5% before the operator took action to correct it. Failure to operate within the procedurally specified limits was a performance deficiency. Technical Specification 5.4.1.a requires, in part, that written procedures be established and implemented covering the procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. 2BGP 100-3, POWER ASCENSION, states in step E.1.e, the rate of Reactor Power rise shall be limited per NF-AP-440 and refers the operator to the ReMA guidance. In step 5.3.4 of NF-AP-440, PWR FUEL CONDITIONING, operators are directed to maintain AO within 3% of target when increasing power above 75% of rated thermal power. The load following ReMA specified a target value of -0.5% for AO. Contrary to the TS 5.4.1.a requirements specified above, the operator did not implement the actions specified in procedures established for changing power and load following to control the key parameter within the specified control band. The operator identified that AO was outside the specified band and the crew immediately took action to restore AO to within the ReMA limits by withdrawing control rods. The issue was entered into CAP as IR 02575960, and the operating department implanted prompt action to communicate the cause of the error to all operators and qualified nuclear engineers. In addition, additional management observations of power maneuvering activities were put in place. The inspectors determined this issue was more than minor because the performance deficiency impacted the Human Performance attribute of the Barrier Integrity Cornerstone and adversely impacted the cornerstone objective to provide reasonable assurance that the physical design barrier (fuel cladding) protect the public from radionuclide releases caused by accidents or events. Specifically, failure to follow fuel conditioning guidelines to monitor and control key parameters while making reactivity changes could result in fuel clad damage and adversely impact nuclear safety. The inspectors determined that the issue was of very low safety significance (Green) because the axial offset was still within the bounding limits established and analyzed by the core operating limit report and no fuel damage occurred.
05000454/FIN-2015008-0330 September 2015 23:59:59ByronNRC identifiedFailure to Evaluate the Adverse Effects of Sharing the RWSTs of Both Reactor UnitsThe team identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, and an associated finding of very-low safety significance (Green) for the licensees failure to perform a written safety evaluation that provided the bases for the determination that a change which resulted in the sharing of the refueling water storage tanks (RWSTs) of both reactor units did not require a license amendment. Specifically, the licensee did not evaluate the adverse effect of reducing reactor unit independence. The licensee captured this issue into their CAP with a proposed action to revise the associated calculation to remove the dependence on the opposite unit, and/or review the implications of crediting the opposite unit RWST under their 10 CFR 50.59 process. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. In addition, it was associated with the Barrier Integrity cornerstone attribute of design control, and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. In addition, the associated traditional enforcement violation was more than minor because the team could not reasonably determine that the changes would not have ultimately required NRC prior approval. The finding screened as very-low safety significance (Green) because it did not result in the loss of operability or functionality, and it did not represent an actual open pathway in the physical integrity of the reactor containment. Specifically, the licensee reviewed the affected calculation and reasonably determined that enough conservatism existed such that adequate net positive suction head (NPSH) could be maintained without sharing the RWSTs of both reactor units. The team did not identify a cross-cutting aspect associated with this finding because it was confirmed not to be reflective of current performance due to the age of the performance deficiency.
05000454/FIN-2015008-0130 September 2015 23:59:59ByronNRC identifiedQuestion Regarding the Maximum Wet Bulb Temperature Value Assumed in the SXCT Tornado AnalysisQuestion Regarding the Maximum Wet Bulb Temperature Value Assumed in the Emergency Service Water Cooling Tower Tornado Analysis Introduction: The team identified an unresolved item (URI) regarding the maximum wet-bulb temperature value assumed in the SXCT tornado analysis. Specifically, the team noted the analysis used a value which was less restrictive than the highest 3-hour wet-bulb temperature recorded for the site as described in the UFSAR. Description: In Section 3.5.4 of the UFSAR, Analysis of Missiles Generated by a Tornado, stated that, An analysis of the UHS cooling capability for a tornado missile event has been made. It also stated that, A maximum outside air wet-bulb temperature of 78 degrees Fahrenheit is assumed and is conservatively held constant throughout the transient. In addition, this UFSAR section stated that, The analysis was performed using service water cooling tower performance curves generated using the method described in UFSAR Section 9.2.5.3.1.1.2 (...). The analysis of the UHS cooling capability for a tornado missile event was calculation BYR09-002, UHS Capability with Loss of SX (Emergency Service Water) Fans due to a Tornado Event, which used a constant maximum outside air wet-bulb temperature value of 78 degrees Fahrenheit consistent with UFSAR Section 3.5.4. However, the team noted the assumed maximum outside air wet-bulb temperature value of 78 degrees Fahrenheit appeared to be inconsistent with the method described in UFSAR Section 9.2.5.3.1.1.2, Steady State Tower Performance Analysis. Specifically, it stated that, The design wet-bulb temperature during warm weather operation is 82 degrees Fahrenheit (Refer to UFSAR Section 2.3.1.2.4). In Section 2.3.1.2.4 of the UFSAR, Ultimate Heat Sink Design, stated that, This analysis (described in Section 9.2.5.3.1.1) includes scenarios with the highest 3-hour wet-bulb temperature, 82 degrees Fahrenheit, which was recorded on July 30, 1961, at 3:00 pm. This UFSAR section also stated that, Per Regulatory Guide 1.27, the ultimate heat sink must be capable of performing its cooling function during the design basis event for this worst case 3-hour wet-bulb temperature. In addition, it stated, However, the design operating wet-bulb temperature of the ultimate heat sink is 78 degrees Fahrenheit (ASHRAE 1 percent exceedance value). This issue is unresolved pending further review by the Office of Nuclear Reactor Regulation (NRR) of the licensing basis related to the wet-bulb temperature value applicable for the SXCT tornado analysis, and the team determination of further NRC actions to resolve the issue. (URI 05000454/2015008-01; 05000455/2015008-01, Question Regarding the Maximum Wet-Bulb Temperature Value Assumed in the SXCT Tornado Analysis)
05000454/FIN-2015008-1130 September 2015 23:59:59ByronNRC identifiedOperability Evaluation Relied on Probabilities of Occurrence of the Associated EventThe team identified a finding of very-low safety significance (Green), and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to make an operability determination without relying on the use of probabilistic tools. Specifically, an operability evaluation for an SXCT degraded condition used probabilities of occurrence of tornado events which was contrary to The requirements of the licensee procedure established for assessing operability of structures, systems, and components (SSCs). The licensee captured the teams concern in their CAP to revise the affected operability evaluation without using probability of occurrence of tornado events. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external events, and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) using a detailed evaluation because a loss of SXCT during a tornado event would degrade one or more trains of a system that supports a risk-significant system or function. The bounding change to the core damage frequency was less than 5.4E-8/year. The team determined that this finding had a cross-cutting aspect in the area of human performance because the licensee did not ensure knowledge transfer to maintain a knowledgeable and technically competent workforce. Specifically, the licensee did not ensure personnel were trained on the prohibition of the use of probabilities of occurrence of an event when performing operability evaluations, which was contained in licensee procedure established for assessing operability of SSCs. (H.9)
05000454/FIN-2015008-1030 September 2015 23:59:59ByronNRC identifiedFailure to Maintain the Instrument Loops Used to Verify Compliance with the Containment Average Air Temperature TS LimitThe team identified a finding of very-low safety significance, and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to have procedures to maintain the accuracy within necessary limits of the instrument loops used to verify compliance with the containment average air temperature TS limit of 120 degrees Fahrenheit. Specifically, in 2007, the licensee cancelled the periodic preventive maintenance (PM) intended to maintain the necessary instrument loops accuracy. The licensee entered this issue into their CAP and reasonably established that the 120 degrees Fahrenheit limit was not exceeded by reviewing applicable historical records from 2002 to time of this inspection. The performance deficiency was determined to be more than minor because it was associated with the configuration control attribute of the Barrier Integrity Cornerstone, and adversely affected the cornerstone objective to ensure that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding screened as very-low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of reactor containment or involved an actual reduction in hydrogen igniter function. Specifically, the containment integrity remained intact and the finding did not impact the hydrogen igniter function. The team determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee did not identify issues completely and accurately in accordance with the CAP. Specifically, on January 15, 2015, the licensee captured the lack of periodic PM activities for the containment air temperature instrument loops in the CAP. However, the licensee failed to completely and accurately identify the issue in that it was not treated as a CAQ. As a consequence, no corrective actions were implemented. (P.1)
05000454/FIN-2015008-0630 September 2015 23:59:59ByronNRC identifiedFailure to Evaluate the Adverse Effects of Changing the SXCT Tornado Analysis as Described in the UFSARThe team identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, and an associated finding of very-low safety significance (Green) for the licensees failure to perform a written evaluation that provided the bases for the determination that the changes to the emergency service water cooling tower (SXCT) tornado analysis as described in the UFSAR did not require a license amendment. Specifically, the associated 10 CFR 50.59 Evaluation did not address the introduction of a new failure mode, the resulting loss of heat removal capacity during worst postulated conditions, and addition of operator actions that have not been demonstrated can be completed within the required time to restore the required SXCT heat removal capacity during worst case conditions. The licensee captured this issue in their CAP with a proposed action to revise the 10 CFR 50.59 Evaluation and submit a Licensee Amendment Request. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external events, and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. In addition, the associated tradition enforcement violation was determined to be more than minor because the team could not reasonably determine that the changes would not have ultimately required prior NRC approval. The finding screened as of very-low safety significance (Green) using a detailed evaluation because a loss of SXCT during a tornado event would degrade one or more trains of a system that supports a risk-significant system or function. The bounding change to the core damage frequency was less than 5.4E-8/year. The team did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance due to the age of the performance deficiency.