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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000237/FIN-2018001-01 | 2018Q1 | Dresden | Enforcement Action: EA18016: Unanalyzed Condition for Tornado Missiles | On June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (ML15020A419), 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, Enforcement Discretion For Tornado Generated Missile Protection Non-Compliance, which was also issued on June 10, 2015 (ML15111A269) and revised on February 7, 2017 (ML16355A286). The discretion applied to Technical Specification (TS) limiting condition for operations (LCOs) that would require a reactor shutdown or mode change if the licensee could not meet the required actions within the TS completion time due to structures, system, and components (SSCs) declared inoperable because of tornado generated missile issues. The EGM stated that a bounding risk analysis performed for this issue concluded that tornado missile scenarios do not represent an immediate safety concern because their risk is within the LIC504, Integrated Risk-Informed Decision-Making Process for Emergent Issues, risk acceptance guidelines. In the case of Dresden Station, the EGM provided for enforcement discretion of up to three years from the original date of issuance of the EGM. The EGM allowed the licensee to re-establish operability when the licensee implemented, prior to the expiration of the time mandated by the affected LCOs, initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened followed by more comprehensive compensatory measures within 60 days of issue discovery. The enforcement discretion was also conditional to the comprehensive measures remaining in place until permanent repairs are completed or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Section 3.5 of the Dresden Power Station Updated Final Safety Analysis Report (UFSAR) states in part that SSCs important to safety shall be adequately protected against missiles generated by various causes, including natural phenomena. On February 12, 2018, the licensee initiated IR 04103159, identifying a nonconforming condition of Section 3.5. Specifically, the vent lines for the U2, U2/3, and U3 emergency diesel generator (EDG) fuel oil tanks were not adequately protected from tornado-generated missiles. The licensee declared fuel oil tanks and their associated EDGs inoperable, and promptly implemented compensatory measures designed to reduce the likelihood of tornado-generated missile effects. The condition was reported to the NRC as Event Notice (EN) 53204 as an unanalyzed condition and potential loss of safety function. Corrective Action(s): The licensee documented the inoperability of the SSCs in the Corrective Action Program (CAP) and in the control room operating log. In addition, the affected TS LCO conditions applicable in the mode of operation at the time of discovery were documented 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 licensees immediate compensatory measures included: Verifying that procedures were in place and training was current for performing actions in response to a tornado event. Verifying that procedures were in place and training was current to respond to a tornado watch, such as: (1) actions to be taken relating to tornado missile hazards; (2) potential restoration of equipment important to maintaining safe shutdown conditions that is unavailable at the time of the tornado watch; (3) warning and protection strategies for personnel; and (4) damage assessment and restorative actions for equipment that may be damaged during a tornado. Establishing a heightened level of station awareness and preparedness relative to identified tornado missile vulnerabilities. The licensees longer term compensatory measure was to modify DOA001002, Tornado Warning Severe Winds procedure to include actions for damage assessment and restorative actions for systems with a vulnerability to damage from tornado missiles. Corrective Action Reference: IR 04103159 |
| 05000298/FIN-2007008-01 | 2007Q4 | Cooper | Inadequate POST-FIRE Safe Shutdown Procedures | Technical Specification, Section 5.4.1.d states, Written Procedures shall be established, implemented and maintained covering the following activities: d. Fire Protection Program Implementation. The fire protection program relies on manual actions performed outside of the control room for achieving and maintaining hot shutdown as documented in the 10 CFR Part 50, Appendix R, Post-Fire Safe and Alternative Shutdown Analysis Report. The post-fire safe shutdown at the Cooper Nuclear Station requires these operations to be performed in accordance with one of two Emergency Procedures 5.4POST-FIRE, Post-Fire Operational Information, or 5.4FIRE-S/D, Fire Induced Shutdown From Outside Control Room. Contrary to the above, the team concluded that, from 1998 until June, 2007, inadequate procedural guidance was provided to allow operators to successfully perform required post-fire safe shutdown manual operations. Specifically, inadequate procedural guidance was provided in the procedures for the manual operation of 10 motor-operated valves from their motor starters as required by the fire protection program. Because the licensee committed to adopting NFPA Standard 805 and changing their fire protection program license basis to comply with 10 CFR 50.48(c), this issue may be subject to enforcement discretion in accordance with the NRC Enforcement Policy. Pending completion of additional analyses to determine whether this finding is less than high safety significance, and thus whether it should be treated as a violation, this issue is being treated as an unresolved item: URI 05000298/2007008-01, Inadequate Post-Fire Safe Shutdown Procedures. (EA-070204) |
| 05000298/FIN-2008007-01 | 2008Q1 | Cooper | Inadequate POST-FIRE Safe Shutdown Procedures | An apparent violation of 10 CFR Part 50, Appendix B, Criterion V, was identified for failure to ensure that some steps contained in Emergency Procedures at Cooper Nuclear Station would work as written. Inspectors identified that steps in Emergency Procedure 5.4POST-FIRE, Post-Fire Operational Information, and Emergency Procedure 5.4FIRE-S/D, Fire Induced Shutdown From Outside Control Room, intended to reposition motor operated valves locally, would not have worked as written because the steps were not appropriate for the configuration of the motor starter circuits. 10 CFR Part 50, Appendix B, Criterion V was not met because these quality related procedures would not work to allow operators to bring the plant to a safe shutdown condition in the event of certain fires. This finding had a cross-cutting aspect in Problem Identification and Resolution, under the Corrective Action Program attribute, because the licensee did not thoroughly evaluate the 2004 NRC violation to address causes and extent of condition. (P.1.c - Evaluations) This finding is of greater than minor safety significance because it impacted the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to external events (such as fire) to prevent undesirable consequences. This finding affected both the procedure quality and protection against external factors (such as fires) attributes of this cornerstone objective. This finding was determined to have a preliminary Greater Than Green safety significance during a Phase 3 evaluation using best-available information. This problem, which has existed since 1997, involves risk factors that were not dependent on specific fire damage. The scenarios of concern involve larger fires in specific areas of the plant which trigger operators to implement fire response procedures to place the plant in a safe shutdown condition. Since some of those actions could not be completed using the procedures as written, this would challenge the operators ability to establish adequate core cooling. Upon identification of this issue, the licensee took immediate compensatory actions to notify operations of the procedural problems, establish a roving fire watch, issue a night order to communicate to all operating crews, and change the procedures. Both emergency procedures have been revised to assure correct valve alignment. Therefore, this finding does not represent a current safety concern |
| 05000313/FIN-2001006-01 | 2001Q2 | Arkansas Nuclear | Fire Zones 98J and 99M DID Not Meet Requirements of 10 CFR Part 50, Appendix R, Section III.G.2 | During an NRC inspection conducted June 11 - 22, 2001, and July 2 - 13, 2001, a violation of NRC requirements was identified. In accordance with the General Statement of Policy and Procedure for NRC Enforcement Actions, NUREG-1600, the violation is listed below: 10 CFR 50.48, Fire protection, Section (b) states, Appendix R to this part establishes fire protection features required to satisfy Criterion 3 of Appendix A to this part with respect to certain generic issues for nuclear power plants licensed to operate before January 1, 1979. ... With respect to all other fire protection features covered by Appendix R, all nuclear power plants licensed to operate before January 1, 1979, must satisfy the applicable requirements of Appendix R to this part, including specifically the requirements of Sections III.G, III.J, and III.O. 10 CFR Part 50, Appendix R, Paragraph III.G.2 states, Except as provided for in paragraph G.3 of this section, where cables or equipment, including associated non-safety circuits that could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, one of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided: a. Separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3-hour rating. Structural steel forming a part of or supporting such fire barriers shall be protected to provide fire resistance equivalent to that required of the barrier; b. Separation of cables and equipment and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustible or fire hazards. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area; or c. Enclosure of cable and equipment and associated non-safety circuits of one redundant train in a fire barrier having a 1-hour rating. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area ... . 10 CFR Part 50, Appendix R, Paragraph III.G.3 states, Alternative or dedicated shutdown capability and its associated circuits, independent of cables, systems or components in the area, room or zone under consideration, should be provided: a. Where the protection of systems whose function is required for hot shutdown does not satisfy the requirement of paragraph G.2 of this section; or ... -2- Contrary to this requirement, the licensee failed to ensure that cables and equipment of redundant trains of systems necessary to achieve and maintain hot shutdown conditions would remain free of fire damage (in the event of a fire) by one of the means specified in 10 CFR Part 50, Appendix R, Paragraph III.G.2, or by alternative means specified in 10 CFR Part 50, Appendix R, Paragraph III.G.3, for Fire Areas 98J and 99M in Arkansas Nuclear One, Unit 1. This violation is associated with a White significance determination process finding (50-313;368/0106-02). |
| 05000313/FIN-2003002-01 | 2003Q1 | Arkansas Nuclear | Deletion of Containment Integrity Controls for Secondary System Containment Penetrations | Severity Level IV. The inspectors identified a noncited violation of 10 CFR 50.59 because the licensee failed to identify that changes made to the Units 1 and 2 Updated Safety Analysis Reports required a license amendment request. These changes removed containment isolation valve controls for secondary system containment penetrations. The licensee initiated corrective action on March 28, 2003, to prepare a license amendment request to obtain NRC approval of the changes to the Updated Safety Analysis Reports This is an item for traditional enforcement because it involves an issue not appropriate for evaluation using the SDP. It involves a violation of 10 CFR 50.59, an issue which impacts NRC oversight ability. The issue is more than minor because it involves a programmatic issue affecting containment controls for all secondary system penetrations. It was considered to be a noncited Severity Level IV violation. Management review determined it was greater than minor because the change should have received NRC review prior to implementation. Redundant containment barrier (system piping) existed and the licensee entered this issue into its corrective action program. |
| 05000313/FIN-2003003-02 | 2003Q2 | Arkansas Nuclear | Adequacy of means to notify populace in EPZ | The inspectors identified a finding related to the adequacy of the means established for notification of members of the populace in the plume exposure emergency planning zone. Because the finding affected the reactor safety emergency preparedness cornerstone objective, the finding is greater than minor. The finding also was determined to have a potential safety significance greater than very low significance because of the potential degradation of the risk significant planning standard 10 CFR 50.47(b)(5), in that less than 100 percent of the population in the emergency planning zone would be alerted as required by the alert notification system design. This is an unresolved item because additional information is required to determine both the compliance aspects and significance of the finding. One element of the unresolved item is the adequacy of the licensees audit program to be able to identify poor performance or adverse trends by the office of Nuclear Planning and Response Programs implementation of the radio program. Lack of an adequate audit program would result in a lack of identification of problems, which would allow a degraded condition to continue to worsen. The licensee did not conduct audits of the Nuclear Planning and Response Programs, which reduced their opportunity to identify any potential concern in implementation of that program. |
| 05000313/FIN-2003004-05 | 2003Q3 | Arkansas Nuclear | Failure to Obtain a License Amendment for Upgrade of the Spent Fuel Area Crane | A noncited violation of 10 CFR 50.59 was identified by the inspectors when the licensee did not submit a license amendment request for a modification to the L-3 spent fuel area crane. The modification, which increased the maximum critical load rating to allow for a different type of spent fuel storage cask to be carried over the control rooms of both units, created the possibility for a malfunction of the L-3 crane that had a different result than previously evaluated. The licensee subsequently submitted a license amendment request for the modification on February 24, 2003 This issue involves traditional enforcement because it involves a violation of 10 CFR 50.59 and is more than minor because there was a reasonable likelihood that the change would require NRC review and approval prior to its implementation. The finding affects the initiating events cornerstone objective attributable to fuel handling equipment performance and has very low safety significance because, after identification of the problem, the licensee did not transfer spent fuel casks until the license amendment was approved. Consequently, the finding is categorized as a Severity Level IV noncited violation in accordance with the NRC Enforcement Policy. |
| 05000315/FIN-2015008-01 | 2015Q4 | Cook | Failure to Ensure the Required Seven Day EDG Fuel Oil Storage | The team identified a finding of very low safety significance (Green), and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, for the failure to translate the emergency diesel generator (EDG) fuel oil storage design basis into applicable procedures and calculations. Specifically, the required 7-day fuel oil supply did not account for the fact that the fuel oil storage tanks (FOSTs) were shared between the two reactor units. The licensee captured this issue in their Corrective Action Program (CAP) as Action Request (AR) 2015-15019 with a proposed action to revise the applicable calculations and procedures to ensure the FOSTs can supply fuel for seven days while accounting for the diesel fuel oil consumption of both reactor units. The performance deficiency was determined to be 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 systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality of the mitigating systems. Specifically, the licensee performed a past operability review and reasonable determined the FOST remained operable because fuel oil volume was maintained greater than the value established by calculation MD-12-DG-004-N. In addition, the availability of a fuel oil low level alarm with an administrative setpoint greater than the value established by this calculation and the expected relatively slow FOST depletion would have reasonably prompted and allowed operators to initiate actions to conserve fuel had an event occurred. The team did not identify a cross-cutting aspect associated with this finding because it was an original design issue; therefore, it was not reflective of current performance. |
| 05000315/FIN-2015008-02 | 2015Q4 | Cook | Failure to Verify the Acceptability of the Surveillance Acceptance Limits for CRID Inverter Operability | The team identified a finding of very low safety significance (Green), and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to verify the adequacy of the surveillance acceptance limits for control room instrumentation distribution (CRID) inverter operability. Specifically, the licensee did not verify the CRID inverter acceptance limits included in the applicable Technical Specification (TS) Surveillance Requirement procedures were adequate to demonstrate CRID operability. The licensee captured this issue in their CAP as AR 2015-14430 and AR 2015-14607, and established a compensatory action to impose more restrictive acceptance limits. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating System cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee reviewed the affected surveillance results for the last 12 months and reasonably determined operability was maintained because the results were within the vendor specifications. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. |
| 05000315/FIN-2015008-03 | 2015Q4 | Cook | Failure to Verify CRID Inverter Capability to Function During Fault Conditions | The team identified a finding of very low safety significance (Green), and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to verify the CRID inverter capability to interrupt faulted conditions on its output during postulated design basis events. Specifically, the licensee did not ensure that the vital inverter was adequately protected from the effects of a fault occurring at the circuit non-safety related loads. The licensee captured this issue in their CAP as AR 2015-14805 and AR 2015-14807, and reasonably determined the installed non-safety related circuit protective devices would be expected to operate and protect the vital inverter during fault clearing conditions on the non-safety related loads powered by the inverter supplied CRID panel bus. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating System cornerstone, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee evaluated the condition for operability and reasonably determined that installed non-safety related circuit protective devices would be expected to operate and protect the vital inverter during fault clearing conditions on the non-safety related loads powered by the inverter supplied CRID panel bus. The team did not identify a cross-cutting aspect associated with this finding because it was not reflective of current performance. |
| 05000315/FIN-2015008-04 | 2015Q4 | Cook | Failure to Consider All Design Basis CCW Passive Failures | The team identified a finding of very low safety significance (Green), and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to verify that the component cooling water (CCW) design was capable of accepting a passive failure as described in the Updated Final Safety Analysis Report (UFSAR). Specifically, the passive failure definition described in the UFSAR was more limiting than the licensee postulated passive failure. The licensee entered this issue into their CAP as AR 2015-15073 with a proposed plan to reconcile the differences between the design basis and plant documentation. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very low safety significance (Green) because it did not represent a loss of function, an actual loss of function of a single TS train or two separate TS safety systems, or an actual loss of function of one or more non-TS trains. Specifically, the licensee performed a historical review of CCW isolation valve leakage and reasonably determined that actual leakage values would have reasonably allowed sufficient response time to provide system makeup to the redundant train. In addition, the licensee performed a historical review of CCW passive failures and did not find an actual loss of function due to a passive failure. The team did not identify a cross-cutting aspect associated with this finding because it was an original design issue; therefore, it was not reflective of current performance. |
| 05000315/FIN-2015008-05 | 2015Q4 | Cook | Failure to Meet Applicable ISI Requirements for All CCW System Portions Within the ASME Code Class 3 Boundary | The team identified a finding of very low safety significance (Green), and an associated NCV of 10 CFR 50.55a, Codes and Standards, for the failure to meet the Inservice Inspection (ISI) requirements for all CCW components within the American Society of Mechanical Engineers (ASME) Code Class 3 boundary. Specifically, the licensee did not apply the applicable ISI requirements to all portions of the CCW system within the system ASME Code Class 3 boundary because this boundary was not appropriately established or justified. The licensee entered this issue into their CAP as AR 2015-15069 and reasonably determined the CCW remained operable. 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. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed a historical system health review and reasonably determined the CCW remained operable because periodic system walkdowns by the system owner and shiftly rounds by operations had not identified significant CCW system leaks. The team did not identify a cross-cutting aspect associated with this finding because it was not reflective of current performance. |
| 05000315/FIN-2015008-06 | 2015Q4 | Cook | failure to Develop Procedures to Provide Starting Air to the EDGs to Recover From a SBO | The team identified a finding of very low safety significance (Green), and an associated NCV of 10 CFR 50.63, Loss of All Alternating Current Power, for the failure to develop procedures to provide starting air to the EDGs to restore emergency alternating current power when recovering from a station blackout (SBO). Specifically, plant procedures did not ensure that there would be sufficient pressure in the EDG air receivers to start an EDG at the end of a 4-hour SBO coping period. In addition, the licensee did not have another proceduralized method of starting an EDG after a 4-hour period. The licensee entered this issue into their CAP as AR 2015-14802 and established an air receiver leak down rate administrative limit that would reasonably preserve sufficient pressure for four hours until the issue is resolved. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very low safety significance (Green). Specifically, a senior risk analyst performed a detailed risk evaluation and determined that the estimated change in core damage frequency was approximately 1.8E8/yr. The team determined that this finding had a cross cutting aspect in the area of problem identification and resolution because the licensee did not systematically and effectively evaluate relevant external operating experience. Specifically, the licensee self-assessment, conducted in preparation to this inspection, reviewed a similar issue identified at a different station and incorrectly concluded that, This issue is not likely to occur at Cook. |
| 05000315/FIN-2015008-07 | 2015Q4 | Cook | Failure to Verify the Station's Capability to Isolate Postulated CCW System Out-Leakagge | The team identified a finding of very low safety significance (Green), and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to verify the CCW design capability to isolate a postulated CCW system out-leakage. Specifically, the CCW isolation valves were not periodically leak tested, and the system design and plant procedures did not include safety related and/or seismic-qualified makeup capabilities. The licensee entered this issue into their CAP as AR 2015-14961, and established temporary procedures and pre-staged equipment to quickly provide system makeup from alternate sources. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed a historical review of isolation valve leakage and reasonably determined that actual leakage values would have reasonably allowed sufficient response time to provide system makeup. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. |
| 05000315/FIN-2018001-01 | 2018Q1 | Cook | Failure of Unit 1 Turbine Driven Auxiliary Feedwater Pump to Reach Rated Speed | A self-revealed finding of very low safety significance with an associated Non-Cited Violation of Technical Specification 5.4 Procedures, occurred on December 21, 2017, when the Unit 1 Turbine-Driven Auxiliary Feedwater Pump failed to reach rated speed during a surveillance. Procedure 12MHP5021056008, Turbine-Driven Auxiliary Feedwater Pump Governor Valve Maintenance, was not appropriate for the circumstances in that direction was not given to check that the governor valve could fully open following maintenance on the governor valve. |
| 05000315/FIN-2018001-02 | 2018Q1 | Cook | Operation of Letdown System Leads to Voiding and Subsequent Relief Valve Lift | The inspectors identified a finding of very low safety significanceand associated Non-Cited Violation of Technical Specification 5.4, Procedures, when the licensee failed to maintain a procedure for operating the letdown system. As a result, a water-hammer occurred which caused a safety-related relief valve to lift, which discharged reactor coolant to the Pressurizer Relief Tank until letdown was isolated |
| 05000315/FIN-2018002-02 | 2018Q2 | Cook | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Enforcement: Violation: License conditions 2.C.(4) (Unit 1) and 2.C.(3)(o) (Unit 2) require implementation of the approved fire protection program. Per the Cook NFPA 805 Fire Protection Program Manual Sections 3.11.2 and 3.11.4, fire seals shall have at least a three hour fire rating. Contrary to the above, on February 6, 2018, the licensee identified multiple fire seals (many of which were between the control rooms and the cable spreading area underneath) that were degraded to the point that they could no longer meet the three hour rating requirement of Sections 3.11.2 and 3.11.4 of the Cook NFPA 805 Fire Protection Program Manual. Specifically, inadequate controls in the fire seal maintenance procedure and unclear guidance for Performance Verification department inspections led to a deterioration in seal quality. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it adversely affected the Protection Against External Factors attribute of the Mitigating Systems cornerstone, whose objective is to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). The inspectors assessed the significance of the finding usingSignificance Determination Process Appendix F and concluded the violation was of very low safety significance (Green).Corrective Action Reference: AR20181208 |
| 05000315/FIN-2018002-03 | 2018Q2 | Cook | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy Violation: Title 10 of the Code of Federal Regulations (10 CFR) 50.47 b(8) requires that licensee emergency plans meet the standard of having adequate emergency facilities. The Cook Plant Emergency Plan states that the Technical Support Center (TSC) (an emergency facility) will be constructed to provide the same degree of radiological habitability as the Control Room under accident conditions. Contrary to the above, from January 24 to 30, 2018, the licensee failed to maintain the TSC as an adequate emergency facility, by installing a portable air conditioning unit in the Shift Managers office which compromised the ability of the TSC ventilation system to fulfill its function of providing the necessary radiological protection for the TSC. Specifically, the exhaust from the portable unit was routed to an existing ventilation duct of the TSC ventilation system, and a panel on one of the ventilation units was opened, exposing the TSC to the turbine building environment. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it adversely affected the Facilities and Equipment attribute of the Emergency Preparedness cornerstone, whose objective is to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors assessed the significance of the finding usingSDP Appendix B and concluded the violation was of very low safety or security significance (Green). Corrective Action Reference: AR20180952 |
| 05000315/FIN-2018002-04 | 2018Q2 | Cook | Licensee-Identified Violation | This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Enforcement: Violation: Title 10 Code of Federal Regulations; Part 20.1501(c) requires that the licensee shall ensure that instruments and equipment used for quantitative radiation measurements are calibrated periodically for the radiation measured. Contrary to the above, between November 2012 and May 2017 the licensee used the liquid scintillation counter for quantitative radiation measurements outside the range of equipment capability and the system calibration. The licensee analyzed the impact on the annual effluent reports and UFSAR limits between 1/8/2013 and 5/3/2017. The licensee entered the violation on the corrective action program. Licensee Identified Non-Cited Violation Significance/Severity Level: Green. The inspectors determined the performance deficiency was more than minor because it adversely affected the Plant Facilities/Equipment and Instrumentation attribute of the Public Radiation Safety Cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. The inspectors assessed the significance of the finding usingSDP Appendix D and concluded the violation was of very low safety or security significance (Green). Corrective Action Reference: AR20174835 |
| 05000315/FIN-2018002-05 | 2018Q2 | Cook | Minor Violation | While there did appear to be a reduction in operational errors being made in the field while manipulating equipment (such as during clearance activities and in performing certain evolutions) the inspectors noted a trend in configuration control issues. Most of these dealt with some kind of operation department interface or coordination with another department. In one case, valves associated with feedwater heater level control were left closed following a project to replace some of the heaters, which contributed to a manual reactor trip due to high moisture-separator drain tank level when starting the plant following the Unit 2 refueling outage. Other examples were Chemistry and Operations department coordination on an non-essential service water (NESW) valve alignment which led to NESW being isolated to generator seal oil cooling during plant startup, poor coordination between Maintenance and Operations which resulted in a containment penetration being left open, a pressure gauge remaining isolated after the Projects department completed the heater drain pump replacements, and the failure to ensure that valve-closure tests were done following the feedwater heater replacements. Another identified trend was in the area of post-maintenance testing (PMT). During the refueling outage on Unit 2, both the NRC and the licensee identified instances of improper PMTs being scheduled for safety-related equipment. Inspectors identified work on an EDG fuel oil transfer pump that did not have an in-service test (IST) scheduled. The licensee identified the lack of a time response test following a motor-driven AFW pump motor replacement, was a repeat issue from the previous outage. The licensee also identified the lack of an IST following a seal replacement on a CCW pump. In each case, the issues were discovered and corrected before equipment was restored to fully operable status. In response to the trend, the licensee reviewed other work on safety-related equipment for the outage to confirm the proper PMTs would be done. No other issues were identified. Finally, early in the observation period, the inspectors noted a trend in procedure quality for maintenance activities on safety-related equipment. There were instances regarding Turbine-Driven Auxiliary Feedwater (TDAFW) pump linkages where better procedure direction could have precluded binding and governor-valve travel issues. Additionally, while replacing a TDAFW governor, a snap ring was inadvertently left out of a coupling due to insufficient procedure detail. Regarding the EDGs, the licensee discovered instructions for assembly of air start check valves did not contain the torque guidance that the vendor drawings stipulated. In response to this trend, the licensee started to perform deliberate reviews of OE before maintenance on some safety-related equipment, to verify maintenance instructions had up-to-date guidance before starting work. No violations or findings were identified by the inspectors. 12 Licensee management acknowledged the issues discussed by the inspectors. |
| 05000315/FIN-2018002-06 | 2018Q2 | Cook | Minor Violation | Technical Specification (TS) 5.4, Procedures, requires that the applicable procedures recommended in Regulatory Guide 1.33 be established, implemented, and maintained. Regulatory Guide 1.33 states that maintenance that could affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with procedures appropriate to the circumstances. Contrary to this requirement, procedure 12EHP4030056218, Automatic Operation of Auxiliary Feedwater Pumps, was not performed as written in the procedure. Specifically, pages were skipped which resulted in the 2CD EDG inadvertently starting during the surveillance. Screening: The issue resulted in momentary loss of the T21C and T21D vital busses until the 2CD EDG reached rated speed and connected to the busses. The reactor was defueled at the time. One train of spent fuel pool cooling was lost for several minutes, but the other train stayed in service and there was no apparent change in spent fuel pool temperature. The issue screened as minor based on the guidance in IMC 0612 Appendix E because there were no safety consequences and there was no transient of any significance. Violation: This failure to comply with TS 5.4 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. |
| 05000316/FIN-2018002-01 | 2018Q2 | Cook | Steam Dump Closure Caused by Human Error | On May 10, 2018, a Green self-revealed finding and associated Non-Cited Violation occurred when licensee personnel caused the Unit 2 steam dump valves to the condenser to close. Specifically, when tuning the controller for the steam dump valves, licensee personnel left the controller in automatic, resulting in the closure of all the steam dump valves. This caused both the steam generator power operated relief valves and a steam generator safety valve to lift. |
| 05000361/FIN-2007008-01 | 2007Q2 | San Onofre | Review Adequacy of Manual Action Completion Times for Control Room Evacuation. | The team concluded that there were two issues to be resolved. One involved whether the conclusions of the licensees safe shutdown analysis were adequate, and the other involved whether the control room evacuation procedure was adequate to meet the requirements of 10 CFR 50, Appendix R, Sections III.G.3 and III.L. Additional information was needed from the licensee to determine the safety significance. This issue was entered into the corrective action program under Action Request 070600585. The team used the guidance in Inspection Procedure 71111.05T to assess whether the manual actions discussed above were reasonable and feasible. This was done by conducting timed walkthroughs of the manual actions specified for the five fire areas selected as samples for this inspection. These walkthroughs were conducted using qualified plant operators. The completion times determined in these walkthroughs, noted above, indicated that actual completion times were expected to be much shorter than the allowable times from the safe shutdown analysis. In each case, the team judged that there was no immediate safety concern because the actual completion times appeared to be adequate to support safe shutdown. Enforcement. Additional information was needed to determine whether there was a violation of 10 CFR 50, Appendix R, Sections III.G.3 and III.L. Specifically, the concern was that the safe shutdown analysis allowed 30 minutes to isolate letdown, start an emergency diesel generator, initiate auxiliary feedwater, and secure steam generator blowdown, without demonstrating that these completion would ensure that the plant remained within the performance parameters specified in Section III.L. Pending review of additional analysis by the licensee, this will be treated as an unresolved item: URI 05000361; 362/2007008-01, Review Adequacy of Manual Action Completion Times for Control Room Evacuation. |
| 05000373/FIN-2015009-01 | 2015Q3 | LaSalle | Use of an Analytical Method to Determine the Core Operating Limits without Prior NRC Approva | The inspectors identified a Severity Level IV NCV of Technical Specification (TS) Section 5.6.5, for using an analytical method that was not previously reviewed and approved by the NRC. Specifically in 2013, the licensee used TRACG04P code to determine the Oscillation Power Range Monitor setpoints prior to NRC approval. The TRACG04P code was reviewed and approved in April 24, 2015. TS Section 5.6.5.b stated, in part that the analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the TS. The licensee entered this finding into their Corrective Action Program (CAP) as IR 02528609 and IR 02528612 to correct the issue. The inspectors determined that this issue was a performance deficiency and because the issue had the potential to affect the NRCs ability to perform its regulatory function, the inspectors evaluated this performance deficiency in accordance with the traditional enforcement process. Using the Enforcement Manual, the inspectors characterized the violation as Severity Level IV because the underlying analytical method required NRC approval prior to use. The inspectors did not assign a cross-cutting aspect to this violation in accordance with IMC 0612, Section 07.03.c. |
| 05000373/FIN-2015009-02 | 2015Q3 | LaSalle | Failure to Preform a Required 50.59 Evaluation | The inspectors identified a Severity Level IV, NCV of Title 10, Code of Federal Regulations (CFR), Part 50.59, Changes, Tests, and Experiments, and an associated finding of very-low safety significance (Green) for the failure to perform and maintain a written evaluation to demonstrate that a calculation revision did not require a license amendment. Specifically, calculation L-003263, Volume Requirements for ADS Back-up Compressed Gas System (Bottle Banks), was revised and resulted in new required time critical operator manual actions, procedure changes, UFSAR changes, and an update to the TS Surveillance Requirements; however, a 10 CFR 50.59 evaluation was not performed. The licensee entered this finding into their CAP as IR 2528988. The inspectors determined this finding was more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control. and affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to account for new required time critical operator manual actions, procedure changes, Updated Final Safety Analysis Report (UFSAR) changes, and an update to the TS Surveillance Requirements. This finding has a cross-cutting aspect in the area of Problem, Identification, and Resolution, in the area of evaluation because the licensee did not thoroughly evaluate the extent of condition of revising the design calculation Specifically, the licensee failed to evaluate revising design calculation L-003263 resulting in time critical operator manual actions, procedure changes, UFSAR changes, and an update to the TS Surveillance Requirements. |
| 05000373/FIN-2017004-01 | 2017Q4 | LaSalle | Complete versus Truncated Shifts on Proficiency Watches | The inspectors identified an unresolved item (URI) related to the adequacy of the shifts for proficiency watches stood by specific reactor operators (ROs). Clarification was requested for whether the 8-hour proficiency watches stood by only these specific ROs, should be considered complete or truncated watches, which may not meet the requirements of 10 CFR 55.53(e). Description: Title 10 CFR 55.53(e) states, in part: To maintain active status, the licensee shall actively perform the functions of an operator or senior operator on a minimum of seven 8hour or five 12hour shifts per calendar quarter. In NUREG 1021, Revision 11, ES-605 further explains that: In accordance with 10 CFR 55.53(e), to maintain an active status, licensed operators are required to maintain their proficiency by actively performing the functions of an operator or senior operator on at least seven 8hour or five 12hour shifts per calendar quarter. This requirement may be completed with a combination of complete 8 and 12hour shifts (in a position appropriately credited for watch-standing proficiency as discussed below) at sites having a mixed-shift schedule, and watches shall not be truncated when the operator satisfies the minimum quarterly requirement (56 hours). Overtime may be credited if the overtime work is in a position appropriately credited for watch-standing proficiency. As documented in AR 04070501, dated November 3, 2017, it has been LaSalle Stations practice to use an individuals normal shift work hours to determine the length of his/her proficiency watch. While the operating shift crews were assigned to 12hour shifts, those licensed ROs assigned to other staff positions at LaSalle normally worked 8 hours per day. LaSalle refers to these individuals as Administrative ROs. Thus, when 14 LaSalles Administrative ROs stood their proficiency watches, they stood 8hour watches, and turned over to another operator to complete the normal 12hour operating shift. As stated in this AR, 8hour shifts minimized the overtime costs to maintain active licenses for these individuals. The Operator Licensing and Training Branch was requested via Regional Office Interaction ROI1725, Clarification of Complete vs. Truncated Shift for Proficiency Watches, because Administrative ROs stood 8hour proficiency watches, while all other operators stood 12hour shifts. Clarification is needed from the Operator Licensing and Training Branch and the Office of the General Counsel to determine if the current practice meets the requirements of 10 CFR 55.53(e) to maintain an operating license in an active status. (URI 050000373/201700401; 050000374/201700401, Complete versus Truncated Shifts on Proficiency Watches) |
| 05000373/FIN-2017004-02 | 2017Q4 | LaSalle | Failure to Establish Brazing Repair Procedures with Appropriate Acceptance Criteria | A finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to establish instructions with acceptance criteria that were appropriate to the circumstances for the brazing repair of the Unit Common Division I diesel generator (DG) starting air system. Specifically, through worker skill of the craft, the use of a heat sink device was relied upon to ensure that the adjacent joint of a brazed connection did not cross a temperature threshold that could have melted or otherwise unacceptably weakened the filler material; however, the procedure used did not contain any quantitative acceptance criteria for the adjacent joint temperature to determine that this important activity had been satisfactorily accomplished. The finding was considered more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, without quantitative acceptance criteria for temperature of the adjacent joints in close proximity of a brazed connection it is possible that joints could be reheated to near the solidus temperature of the filler material, resulting in joint weakening and potential failure. The licensee entered the issue into its CAP as AR 04090775. Corrective actions included revising procedures associated with brazing repairs to include a temperature value as a quantitative acceptance criteria for determining that important activities have been satisfactorily accomplished and to address the physical condition of the adjacent joint by verifying its conditions under work order (WO) 4702099 performance. The inspectors determined that the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Attachment 0609.04, Initial Characterization of Findings, dated October 7, 2016. Because the finding impacted the Mitigating Systems Cornerstone the inspectors screened the finding through IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012. The finding screened as very low safety significance (Green) because it did not result in the loss of operability or functionality; thus, the inspectors answered No to all of the mitigating system screening questions. The inspectors determined that the finding had a cross-cutting aspect in the area of Human Performance, under the aspect of Work Management. Specifically, WO 4702099 designated DG air start system repair activities as non-code when an American Society of Mechanical Engineers (ASME) code brazing procedure specification, (BPS) 107107BR Revision 0, was being used to satisfy the standard of record, the diesel engine manufacturers standards. (H.5) |
| 05000373/FIN-2017004-03 | 2017Q4 | LaSalle | Primary Containment Structure, Suppression Pool Columns, Downcomer Vent and Downcomer Vent Bracing Did Not Meet Seismic Category I Requirements | A finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to ensure the adequacy of the design for the primary containment, suppression pool columns, downcomer and downcomer vent bracing. Specifically, the inspectors identified three representative examples where the licensee failed to perform adequate design calculations resulting in the design not being in conformance with Seismic Category I requirements as defined in Updated Final Safety Analysis Report (UFSAR) Sections 3.8.1.4.1, 3.8.1.5 and 3.8.6. The licensee documented these violation examples in ARs 4070065, 4074674 and 4070067 and initiated actions to restore compliance. 4 The inspectors determined the licensees failure to perform adequate evaluations to demonstrate Seismic Category I compliance for the primary containment structure, suppression pool columns, downcomer vents and downcomer vent bracing was contrary to the design control measures per 10 CFR Part 50, Appendix B, requirements and was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of design control and adversely affected the Cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The inspectors determined the finding could be evaluated using the SDP in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, Exhibit 3, Barrier Integrity Screening Questions, for the Barrier Integrity Cornerstone (r\eactor containment). The inspector answered no to the Barrier Integrity questions for reactor containment. The finding screened as having very low safety significance (Green). The inspectors determined there was no cross-cutting aspect associated with this finding because the deficiency was a legacy design calculational issue and, therefore, was not indicative of licensees current performance. |
| 05000373/FIN-2017004-04 | 2017Q4 | LaSalle | Failure of Offsite Power Backfeed Procedure to be Appropriate to the Circumstances Caused Unit 1 Scram | A finding of very low safety significance and an associated NCV of LaSalle Technical Specification (TS) 5.4.1, Procedures, occurred on February 13, 2017, for the stations failure to maintain instructions of a type appropriate to the circumstances for energizing offsite electrical systems during a Unit 2 backfeed evolution (an activity affecting quality per Regulatory Guide 1.33). Specifically, the steps of backfeed procedure, LOPAP01, Revision 35, led to a Unit 1 scram because the prescribed switchyard configuration left both units connected to the 345 kilovolt (kv) ring bus, leaving the operating unit susceptible to the large in-rush current induced by the backfeed energization of the Unit 2 main power transformer. As a corrective action from Action Request (AR) 03973724, the licensee revised the backfeed procedure to eliminate the tie between the units on the ring bus when main power transformers are energized. This performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Initiating Events Cornerstone, and adversely affected the Cornerstone objective of limiting the likelihood of events that upset plant stability because it resulted in a Unit 1 Scram. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At Power, issued June 19, 2012, the inspectors determined that this finding was of very low safety significance because, although the performance deficiency caused a reactor scram, it did not result in the loss of mitigation equipment relied upon to transition the plant from the onset of the scram to a stable shutdown condition. The inspectors determined there was no cross-cutting aspect because the performance deficiency was not indicative of licensees current performance since the design modification occurred greater than 3 years before the event. This inspection report will also bring to closure the associated Licensee Event Report, (LER) 05000373/201700300. |
| 05000382/FIN-2000008-04 | 2000Q3 | Waterford | Failure to identify an unreviewed safety question involving resequencing of nonsafety loads to 1E bus | The inspectors determined that the licensee's failure to obtain Commission approval for an Updated Final Safety Analysis Report correction regarding the resequencing of nonsafety loads to a Class 1E bus following a diesel generator start was a violation of 10 CFR 50.59 and constituted an unreviewed safety question. However, it was determined that this issue would not be a violation under the revised 10 CFR 50.59 rule, currently scheduled to be effective January 2001. This judgement is based on the conclusion that the change did not represent more than a minimal increase in the probability of a malfunction of equipment important to safety. Therefore, in accordance with Section 8.1.3 of the NRC Enforcement Manual (NUGEG/BR-0195, Revision 3), enforcement discretion is being exercised after consultation with the Office of Enforcement pursuant to Section VII.B.6 of the NRC Enforcement Policy and a violation is not being issued (EA-99-220). The licensee placed this issue in their corrective action program as Condition Report 98-0763. |
| 05000382/FIN-2006010-01 | 2006Q3 | Waterford | Protection of Safe Shutdown Capabilities | The team identified an apparent violation of License Condition 2.C.9, "Fire Protection (Section 9.5.1, SSER 8)," for failure to ensure that redundant trains of safe shutdown systems in the same fire area were free of fire damage. The licensee credited unapproved manual actions to mitigate the effects of fire damage in lieu of providing physical protection consistent with the technical requirements of 10 CFR Part 50, Appendix R, Section III.G.2. The team considered the manual actions to be reasonable, therefore, the finding was determined to be of very low safety significance. License Condition 2.C. 9 states, "EOI shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility thru Amendment 36 and as approved in the SER through Supplements 9, subject to the following provision: EOI may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdowns in the event of a fire." In NUREG-0787, "Safety Evaluation Report Related to the Operation of Waterford Steam Electric Station, Unit No. 3," Supplement No. 5, the NRC staff evaluation of Safe Shutdown Capability noted, "In SSER No. 3, the staff reported that the one of the redundant trains needed for safe shutdown would be kept free of fire damage by providing separation, fire protection (fire detection, suppression, fire barriers), repairs for cold shutdown equipment, and/or an alternate shutdown capability." The NRC staff also stated in the conclusion of SSER N0. 5, "The technical requirements set forth in Appendix R, as well as the criteria of BTP 9.5-1, have been used as guidelines in the fire protection evaluation above. By letter dated November 10, 1981, the applicant committed to meet the technical requirements of Appendix R to 10 CFR Part 50. The staff evaluated this commitment, along with specific commitments described in this SSER. Subsequently, the staff concluded that the fire protection program, with the accepted deviations listed below is in conformance with the guidelines of Appendix A to BTP 9.5-1, the requirements of Appendix R, and GDC 3, and is, therefore, acceptable." At the time of the inspection, the fire protection program relied on manual actions for fires outside of the control room for achieving and maintaining hot shutdown as documented in calculation EC-F00-026, "Appendix R Revalidation Project Post Fire Safe Shutdown," and Procedure OP-901-524, "Fire in Areas Affecting Safe Shutdown." An example is establishing a charging flow path for the case of a fire in Fire Area RAB 8 (Switchgear Room). If a fire occurred in Fire Zone RAB 8B (Train B Switchgear Room), operators are required to manually close valve CVC-183 in Fire Area RAB 31 to isolate the Volume Control Tank and open breaker CVCEBKRAB 38 in Fire Zone RAB 8C to fail air operated valve CVC-209 (Charging Header Isolation) in the required open position. |
| 05000382/FIN-2006010-02 | 2006Q3 | Waterford | Potential for Loss of Both Trains of Safe Shutdown Equipment From Damage Due to Fire | On June 12, 2006, the licensee determined, while reviewing a preliminary analysis of the feasibility of manual actions designated to be performed in Fire Area RAB-7 (relay room), that the manual actions were not feasible. This was due to postulated environmental conditions (smoke) that could be present in the fire area during a fire in any adjacent zone of the fire area. The preliminary analysis indicated that the manual actions could not be performed within the time prescribed in the licensees post-fire safe shutdown analysis. Furthermore, the preliminary analysis indicated that there could be damage to equipment in the room needed to achieve safe shutdowns, due to a hot gas layer in the room. Fire Area RAB 7 is divided into four fire zones by partial height walls. These partial height walls were approved in a deviation granted by the NRC in 1984. The licensee determined from fire modeling that a manual action required in 10 minutes may not be feasible with the amount of smoke in the area. The licensee's immediate corrective action was to initiate a fire watch in the fire area. However, since a continuous fire watch was already in place in Fire Area RAB 7, no additional action was needed. This issue was entered into the licensee's corrective action program as Condition Report CR-WF3-2006-01735. Subsequently, the licensee determined that the manual action described in Procedure OP-501-524 was not required since it could be performed concurrently from the control room. The concern identified by the licensee in a preliminary analysis regarding potential damage to equipment due to a hot gas layer present throughout the room will be evaluated by the licensee in their transition to NFPA Standard 805. This LER was reviewed and no findings of significance were identified. Based upon the interim enforcement policy, enforcement discretion would be granted for violations identified by the licensee in their evaluation of equipment damage due to a hot gas layer in Fire Area RAB 7 since the licensee is in transition to NFPA Standard 805. The teams review concluded that this finding meets the criteria for enforcement discretion for plants in transition to a risk-informed, performance-based fire protection program as allowed per 10 CFR 50.48(c). Since all the criteria were met, the NRC is exercising enforcement discretion for this issue. This LER is closed. |
| 05000416/FIN-2007008-02 | 2007Q4 | Grand Gulf | Verify Continued Operability of RHR Heat Exchanger B Due to Fouling | The team was concerned that the licensee had not taken action to confirm that the thermal performance of RHR Heat Exchanger B remained adequate to remove worst case design basis heat loads. The projected fouling rate was based on limited data, some of which may not have been sufficiently accurate to rely on over a long period. Also, the licensees historical data was not sufficient to provide high confidence that the fouling created a linear or predictable impact on heat transfer. Therefore, an unresolved item is being issued to further assess the capability of RHR Heat Exchanger B and determine whether a performance deficiency exists. In response to this concern, the licensee stated their intent to clean and/or conduct a thermal performance test of this heat exchanger prior to the onset of warm weather to ensure that the this component remained capable of removing the design basis heat load. The inspectors will review the results of the testing and/or cleaning when it is completed. Additional information was needed to determine whether there were any violations of NRC requirements associated with this issue. This issue will be tracked as an unresolved item to verify that the fouling did not involve a loss of function: URI 05000416/2007008-02, Verify continued operability of RHR Heat Exchanger B due to fouling. |
| 05000445/FIN-2008006-03 | 2008Q2 | Comanche Peak | Inadequate Fire Suppression Systems | A noncited violation of Unit 1, License Condition 2.G, Fire Protection, was identified for the fire suppression systems in Fire Zones SE16 and SE18 (remote safety related panels/Train B switchgear rooms) not being installed in accordance with the approved fire protection program. The fire suppression systems in Fire Zones SE16 and SE18 are manually actuated dry pipe deluge (pre-action) systems with closed sprinkler heads. The actual configuration did not provide protection in the areas containing one train of safe shutdown cables enclosed in 1-hour fire barriers. The team determined that the fire suppression systems in Fire Zones SE16 and SE18 were not installed in accordance with the configurations in Calculation 0210-63-0064, Partial Sprinkler Coverage Evaluation. The configurations in this calculation were approved by the NRC as the basis for allowing suppression systems with less than full area coverage. The configuration also did not meet the National Fire Protection Association codes. The licensee entered this finding into its corrective action program under Smart Form SMF-2008-000324-00. Failure to ensure the installed fire suppression systems met the requirements of the approved fire protection program was a performance deficiency. This finding was more than minor because it is associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone and could affect the availability, reliability, and capability of systems that respond to fire events to prevent undesirable consequences. The significance of this finding was assessed using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. In completing the Fire Protection Significance Determination Process, Phase 1 and 2 worksheets, it was determined that no potential ignition source could potentially have a direct impact on the cable raceways protected by fire barriers or their supports and that the largest potential ignition sources in the fire zones could not form a hot gas layer sufficient to impact the protected cable raceways or their supports. The evaluation indicated that the finding had a very low safety significance (Green) during the Phase 2 significance determination process |
| 05000445/FIN-2008006-05 | 2008Q2 | Comanche Peak | Inadequate POST-FIRE Safe Shutdown Procedures | A noncited violation of Technical Specification 5.4.1.d was identified concerning the failure to maintain adequate written procedures covering fire protection program implementation. Specifically, procedures for operation of Valves 1-8000A and 1-8000B (power-operated relief valve block valves) and Valves 1-8701A and 1-8702B (residual heat removal loop hot-leg recirculation valves) had local manual actions that might not be completed successfully because of potential fire damage. Procedures ABN-804A, Response to a Fire in the Safeguards Building, Revision 5, and ABN-806A, Response to a Fire in the Electrical and Control Building, Revision 5, directed operators to open the valves from their electrical power supplies because of potential fire damage to control circuits between the main control room and the electrical breakers. Plant operators were instructed to depress a breaker contactor to stroke the valve open. After the operator depresses the contactor, control power is required to hold the contactor closed while the valve strokes. The team identified that potential fire damage to control circuits between the main control room and the electrical breakers could cause a control power fuse to fail, preventing the valve from stroking. The licensee has entered this issue into their corrective action program as Smart Form SMF-2008-000311-00. Failure to provide adequate procedures for the implementation of the fire protection program was a performance deficiency. This finding was more than minor because it is associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone and could affect the availability, reliability, and capability of systems that respond to fire events to prevent undesirable consequences. The significance of this finding was assessed using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. The evaluation determined that the procedural deficiency only affected valves required to reach and maintain cold shutdown conditions; therefore, the finding screened as having very low safety significance (Green) |
| 05000454/FIN-2013004-01 | 2013Q3 | Byron | 10 CFR 50.59 Evaluation Affecting Tornado Analysis | During inspection activities associated with the review of tornado protection of the UHS, the inspectors identified an Unresolved Item associated with the implementation of Engineering Change (EC) 385829, Tornado Missile Design Basis for the Essential Service Water Cooling Towers, that potentially departed from an approved method of evaluation described in the FSAR (as updated). In response to 2007 NCV 05000454/2007004-003; 05000455/2007004- 003, Discrepancies with Tornado Analysis, the licensee completed EC 385829, Tornado Missile Design Basis for the Essential Service Water Cooling Towers, dated February 9, 2012. This analysis reviewed the design standards and design requirements for tornado protection of structures, systems, or components (SSCs). The licensee concluded that deficiencies identified in the violation could be addressed analytically; therefore, no physical changes to the system would be required. The licensee adopted an approach that required two of the eight UHS fans to be available for use following a tornado. The inspectors questioned this change as it differed from the description found in the original safety evaluation. Specifically, Section 9.2.5, Ultimate Heat Sink, of Safety Evaluation Report Related to the Operation of Byron Station, Units 1 and 2, dated February 1982, contained the following information regarding tornado protection of the UHS: The UHS function is not affected by tornado missiles as discussed further in Section 3.5.2 of this SER (Safety Evaluation Report). Thus, the requirements of GDC 2 and the guidelines of Regulatory Guides 1.27, Position C.2, regarding UHS protection against natural phenomena, and 1.29 are met. Also, Section 3.5.2, Structures, Systems, and Components to be Protected from Externally Generated Missiles, of Safety Evaluation Report Related to the Operation of Byron Station, Units 1 and 2, dated February 1982, stated, in part: Each unit has one essential service water cooling tower composed of four cells which serves as the ultimate heat sink. The towers are concrete structures designed to withstand tornado-missile impact. However, exposed piping on the towers, and the cooling tower fans and fan motor drives located on top of the towers are not protected from tornado-generated missiles. The applicant has committed (Tramm letter dated January 2, 1982) to provide protection for all piping external to the missile-proof cooling tower walls. The applicant has also provided the results of an analysis (Tramm letter dated January 2, 1982) which shows that in the event of a failure of all the essential service water cooling tower fans, the essential service water system temperature can be maintained within acceptable limits for proper operation of safety-related equipment served in both units with the towers functioning strictly in a natural draft cooling mode and makeup available from the tornado-missile protected onsite wells. The staff concurs that the applicant has satisfactorily demonstrated the availability of the essential service water system and ultimate heat sink in the event of postulated tornado missiles. Thus, the design of the essential service water cooling towers meets the guidelines of Regulatory Guide 1.27, Ultimate Heat Sink for Nuclear Power Plants, with respect to missile protection. The inspectors also identified information in UFSAR Section 3.5.4.1 that indicated that the UHS was expected to be able to meet its design requirements without operating fans. The following first appeared in a January 2, 1982, licensee response to questions by NRC staff regarding UHS performance. These words, or similar words, continued to appear through Revision 13 of the UFSAR. An analysis of cooling tower capacity without fans has been made. Using the most conservative design conditions, it is predicted if the plant is shut down under non-LOCA conditions with loss of offsite power, the temperature of the service water supplied to the plant will not exceed 110F. Although this exceeds the normal maximum temperature of 100F, no adverse impact on safety equipment will result. In December 2012, the UFSAR was updated to Revision 14. The following language in Revision 14 was selected to be in alignment with the methodology outlined in EC 385829. An analysis of the UHS cooling capability for a tornado missile event has been made. 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 and the time dependent two cooling tower model described in UFSAR Section 9.2.5.3.1.1.3. The following inputs and assumptions were used in the analysis: a. A single tornado generated missile is assumed to disable two essential service water cooling tower (SXCT) fans. Concurrent with the missile impact, a LOOP (Loss of Offsite Power) and an electrical failure is assumed to occur that results in the loss of power to two additional SXCT fans. Additionally up to two SXCT fans are assumed to be initially out of service... The inspectors determined that the method of analysis used in the original safety analysis only relied on equipment that was protected from a tornado to evaluate the designs capability to mitigate the consequences of a tornado. As the fans, fan motors, and fan drives were not protected, they were not relied upon. As stated in the original SER, the design complied with the GDCs, due to the designs ability to maintain service water temperatures within design limits while operating strictly in a natural draft cooling mode. The inspectors questioned whether the licensees revised approach which credited equipment that was not protected from a tornado to mitigate the consequences of a tornado constituted a change in a method of evaluation as described in the UFSAR and associated safety analyses. The licensee entered this issue into their CAP as IR 1546621, Inadequate 50.59 for EC 385829 (SXCT Tornado Missiles). The inspectors have engaged the Office of Nuclear Reactor Regulation (NRR) for clarification regarding the licensees current licensing basis. This issue is an Unresolved Item (URI 05000454/2013004-001; 05000455/2013004-001) pending a response from NRR and the completion of additional reviews. |
| 05000454/FIN-2015004-01 | 2015Q4 | Byron | Inaccurate Technical Basis for Operability Evaluation of Reactor Head Flange Damage | The 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-2015004-02 | 2015Q4 | Byron | Mispositioned Valve in Diesel Fuel Oil Transfer Pump Recirculation Flow Path | A 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-2015004-03 | 2015Q4 | Byron | Failure to Implement Protective Tagging Procedure Requirements | A 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). |
| 05000454/FIN-2015004-04 | 2015Q4 | Byron | Licensee-Identified Violation | On 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-2015004-06 | 2015Q4 | Byron | Licensee-Identified Violation | As 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-2015004-07 | 2015Q4 | Byron | Licensee-Identified Violation | The 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-2015007-04 | 2015Q3 | Byron | Potentially Inadequate Evaluation/Corrective Actions: Diesel Oil Storage Tank (DOST) Vent Line Seismic Supports and Tornado Missile Protection | The inspectors identified a concern that the licensees evaluation and corrective actions may have been inadequate following identification that DOST vent lines may not be seismically supported and adequately protected against tornado missiles as described in the Updated Final Safety Analysis Report (UFSAR) and in an NRC Safety Evaluation Report (SER). The deficient condition was NRC-identified and documented as an NCV in NRC Inspection Report 05000454/2009004-02; 05000455/2009004-02, which was issued in 2009; and was entered into the licensee CAP as AR 877430 and AR 933712. According to the description in UFSAR Section 9.5.4.2, although the DOSTs are safety-related tanks required to remain operable during a tornado event, the tank fill and vent lines are classified as nonsafety-related. In response to an NRC reviewer comment during the license application process that these lines should be designed safety-related and tornado missile protected, the licensee provided a formal response which included, in part, the following (Question 040.99, Commonwealth Edison letter from T. R. Tramm to Harold R. Denton, dated December 28, 1981): Additional supports will be added to maintain integrity of the lines during design basis seismic events; Fill and vent lines are not safety-related while the tanks are; Impact from a tornado missile will not result in loss of function as breakage will occur before crimping; In the event of damage, a capped off 4 Category I line could be opened for use as an emergency vent or fill line; and A 4 Category I tank overflow line could be used as a vent. This position was accepted by the NRC based on the following as documented in the SER (NUREG 0876, February 1982, Section 9.5.4.2): Commitment to seismically support the fill and vent lines; In case of damage due to tornado missiles, availability of unused flanged connections that can be used as fill and vent openings; and the Lines are designed to American National Standards Institute (ANSI) B31.1 Following a review of available licensee documentation, the inspectors identified that the commitment for seismically supporting the vent lines may not have been met. The availability of the unused flanged connections was also uncertain as there were no administrative procedures in place for such actions. The licensee was also unable to justify the break prior to leak statement. Instead, the licensee was relying on the vent path through the overflow line as was discussed in the response to Question 040.99 above. Because of the loop seals in the overflow piping, additional evaluations and loop seal modifications were performed to demonstrate that the tanks were structurally adequate for the maximum vacuum associated with the loop seal configurations. The code of construction for the tanks was the American Society of Mechanical Engineers (ASME) Section III, Division 1, sub-section ND, 1974 edition. However, due to a partial vacuum resulting from the use of overflow piping as a vent path, the use of a methodology that was different than that described in licensee calculation BYR13096 was required. Based on discussions with the licensee, the licensees view is that since the alternate vent path is provided, the vent lines no longer require seismic supports. The inspectors interpretation based on the SER, UFSAR, and the licensee response to the NRC question as described above is that the acceptance of the alternate vent path scenario was applicable to the tornado missile event only, and that the licensee was still required to seismically support the vent lines. Additionally, while the licensee response to Question 040.99 included a discussion of using overflow lines as alternate vent paths, the NRC acceptance was based on the use of unused flanged lines. Further review is needed to determine the correct interpretation of the SER description and the design basis for seismic support and tornado missile protection requirements for the DOST vent lines. This issue will remain open pending further NRC review to ensure that the licensee is in compliance with their current licensing basis. (URI 05000454/2015007-04; 05000455/2015007-04, Potentially Inadequate Evaluation/Corrective Actions: Diesel Oil Storage Tank (DOST) Vent Line Seismic Supports and Tornado Missile Protection) |
| 05000454/FIN-2016002-01 | 2016Q2 | Byron | Failure to Perform ASME Code Case Required Extent of Condition to Identify Unacceptable Piping Flaws | A 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-2016002-02 | 2016Q2 | Byron | Failure to Comply With Radiation Work Permit Requirements Resulting In An Unplanned Dose Rate Alarm | A 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-2016002-03 | 2016Q2 | Byron | Licensee-Identified Violation | Title 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-2016002-04 | 2016Q2 | Byron | Licensee-Identified Violation | Technical 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-2016002-05 | 2016Q2 | Byron | Licensee Implementation of Enforcement Guidance Memorandum 15002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance | On 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. |
| 05000455/FIN-2015004-05 | 2015Q4 | Byron | Licensee-Identified Violation | On 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. |
| 05000461/FIN-2016002-01 | 2016Q2 | Clinton | Material Unsecure in the Secured Material Zone | The inspectors identified a finding of very low safety significance for the failure to ensure material placed within the transformer secured material zone, was secured as required by station procedure MAAA716026, Station Housekeeping/Material Condition Program, Revision 14. Specifically, the inspectors identified unsecured scaffold poles and knuckles within the licensee established secure material zone. The licensee has entered this issue into their corrective action program (CAP) as action request AR 02668245. The material was immediately removed from the secured zone by the licensee. The inspectors determined the licensees failure to ensure material within the secured material zone was adequately secured in accordance with procedure MAAA716026, Station Housekeeping/Material Condition Program, was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations, and is therefore a finding. Specifically, by not securing material in the vicinity of main power transformers, the material could become a missile and impact the transformers causing a potential reactor SCRAM. The finding was screened against the Initiating Events Cornerstone and determined to be of very low safety significance (Green) because the finding did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause an initiating event and did not affect mitigation equipment. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance in the aspect of field presence, where leaders are commonly seen in the work areas of the plant observing, coaching, and reinforcing standards and expectations. Specifically, since initial identification of the issue the inspectors have noted that while discussions on when to perform walkdowns took place, the supervisors or managers did not ensure sufficient field presence to reinforce the standards and expectations, leading to material continuing to be easily found by the inspectors. (H.2) |