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05000461/FIN-2018002-022018Q2ClintonFailure to Establish Adequate Leak Rate Test Procedures for Shutdown Service Water Isolation Valve TestingThe inspectors identified a Green finding and a Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to ensure testing of the shutdown service water (SX) isolation valves was performed with procedures which: (1) incorporated the requirements and acceptance limits contained in applicable design documents; and (2) included provisions for assuring that all prerequisites for the given test had been met. Specifically, the licensee failed to establish leak rate test procedures for SX boundary valves 1CC075A and 1CC076A that included provisions for ensuring the required differential test pressure was met during testing.
05000461/FIN-2018002-012018Q2ClintonFailure to Perform an Operability Determination for Suspected Leakage Past Shutdown Service Water Isolation ValvesThe inspectors identified a Green finding for the failure to perform an operability determination in accordance with Procedure OPAA108115, Operability Determinations (CM1). Specifically, the licensee failed to determine and document the operability status of the shutdown service water system and the ultimate heat sink after the discovery of leakage past the 1CC075A and 1CC076A isolation valves.
05000461/FIN-2018002-052018Q2ClintonMinor ViolationThe inspectors reviewed AR 4116223, Blown Fuses during CPS 9080.23 8.4 for Fast Transfers. The inspectors selected this sample for review due to repetitive fuse failures within the safety-related Division 3 NUS Modules dating back to 2013. As appropriate, the inspectors verified the following attributes during their review: complete and accurate identification of the problem in a timely manner commensurate with its safety significance and ease of discovery; consideration of the extent of condition, generic implications, common cause, and previous occurrences; evaluation and disposition of operability/functionality/reportability issues; classification and prioritization of the resolution of the problem commensurate with safety significance; identification of corrective actions, which were appropriately focused to correct the problem; and completion of corrective actions in a timely manner commensurate with the safety significance of the issue. Description: While reviewing the historical ARs associated with the NUS fuse failures, the inspectors discovered licensee information indicating the NUS fuse failures were likely caused by voltage/current transients within the upstream, safety-related 480V to 120V regulating transformer. The purpose of the transformer was to regulate voltage and current to the downstream components including the NUS modules. However, degradation in the transformers ability to regulate voltage and current levels could create a condition where the voltage and current levels exceeded the NUS fuse rating causing fuse failure. The licensee documented the potential transformer degradation issue on September 20, 2013, in AR 1561455, Division 3, Group 1 Instruments Found De-energized during CPS 9080.23, Specifically, the licensee stated, The most probable cause of the failure of the NUS modules was the transient voltage overshoot of the regulating transformer causing the transient protection varistors on the five NUS modules to actuate, drawing a near fault current until the individual and line feed fuses blew. Station procedure PI-AA-125, Corrective Action Program, defined equipment failure as, damage to or degradation of a system, structure or component that may cause or contribute to the event. Based on the information documented in AR 1561455, the licensee identified transient voltage overshoots in the 480V to 120V regulating transformer, which was a degraded condition causing the NUS modules to fail. Per the licensee definition this would constitute an equipment failure. No further action was taken to identify and correct the regulating transformer degradation until the transformer failed on March 18, 2018, impacting multiple pieces of safety-related Division 3 equipment. Minor Violation: Title 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, requires conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to this requirement, on September 20, 2013, the licensee identified a failure of the 480V to 120V regulating transformer, which manifested itself as a voltage overshoot causing the failure of the NUS modules, but failed to take actions to correct the condition. On March 18, 2018, the regulating transformer subsequently degraded further causing it to fail in a manner that tripped the upstream breaker and impacted additional pieces of safety-related Division 3 equipment. Screening: This issue screened as minor because all the questions associated with a minor issue found in IMC 0612, Appendix B, were answered No. Specifically, the inspectors determined that although the transformer failure affected Division 3 equipment, the failure would not have impacted the Division 3 equipments ability to respond to a DBE or the capability to shut down the reactor and maintain it in a safe shutdown condition. Enforcement: The failure to comply with 10 CFR 50, Appendix B, Criterion XVI, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000461/FIN-2018002-042018Q2ClintonMinor ViolationThe inspectors reviewed AR 4082490, Reactor SCRAM from Trip of 1AP07EJ. The inspectors selected this sample for review due to the safety significance of the Division 1 and 2 safety-related transformers, which is the subject of the AR. This review focused on actions associated with newly installed Divisions 1 and 2 4160V to 480V transformers. As appropriate, the inspectors verified the following attributes during their review of the licensee's corrective actions for the above condition report and other related condition reports: classification and prioritization of the resolution of the problem commensurate with safety significance; and completion of corrective actions in a timely manner commensurate with the safety significance of the issue. The inspectors discussed the corrective actions and associated evaluations with licensee personnel. As a result of this review the inspectors identified the following minor violation: Minor Violation: The inspectors identified a violation of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to follow procedures associated with the CAP. Specifically, on May 10, 2018, the licensee identified discrepant results while testing safety-related transformers 0AP06E2 and 1AP12E2 but failed to enter this issue into the CAP in accordance with PIAA120, Issue Identification and Screening Process, Revision 8, Step 4.3.4, until prompted by the inspectors. Instead, the licensee evaluated the discrepant results within the work order and found them to be acceptable. The licensee generated AR 4137994, Insulation Power Factor Results For 0AP06E & 1AP12E, dated May 15, 2018, after being challenged by the inspectors regarding the need to enter the discrepant test results into the CAP. Screening: This issue screened as minor because all the questions associated with a minor issue found in IMC 0612, Appendix B, were answered No. The failure to document the discrepant values in the CAP did not adversely impact the safety-related transformers. Enforcement: This failure to comply with 10 CFR 50, Appendix B, Criterion II, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. Enforcement: The inspectors did not identify a violation of regulatory requirements associated with this minor finding because the procedure the licensee failed to follow was a self-imposed standard.
05000461/FIN-2018002-032018Q2ClintonMinor ViolationDuring the inspection quarter, the inspectors reviewed a significant number of licensee CAP documents to assess the following performance attributes: complete, accurate, and timely documentation of the identified problem in the CAP; evaluation and timely disposition of operability and reportability issues; consideration of extent of condition and cause, generic implications, common cause, and previous occurrences; classification and prioritization of the problems resolution commensurate with the safety significance; and identification of negative trends associated with human or equipment performance that can potentially impact nuclear safety. Minor Performance Deficiency: The inspectors determined that issues which could impact the operability of TS-related equipment were generally entered into the CAP in a timely manner. However, operability determinations were not always performed within the timeframes established in Section 4.1 of Procedure OPAA108115, Operability Determinations (CM1), because some issue reports were not directly routed to the operating shift crew for review. The CAP software program used by the licensee included a standard set of questions which were normally answered by the individual entering the issue into the CAP. Depending on the answers to the questions, the CAP document routing could automatically bypass the operating shift crew for review. Screening: This issue screened as minor because all the questions associated with a minor issue found in IMC 0612, Appendix B, were answered No. The inspectors did not identify any instance where the failure to perform a timely operability determination had a significant consequence on licensed activities. However, the inspectors discussed the vulnerability between the CAP and the operability determination process with the licensee. The licensee implemented a standing order to require a shift review by the operating crew of condition reports not directly routed to the shift. In addition, the licensee is trending the number of condition reports which are returned by the Station Ownership Committee to the shift for review to determine whether further actions are warranted. Enforcement: The inspectors did not identify a violation of regulatory requirements associated with this minor finding because the procedure the licensee failed to follow was a self-imposed standard.
05000254/FIN-2018002-012018Q2Quad CitiesFailure to Have a Procedure Appropriate to Circumstances for Degraded Voltage RelaysA finding of very low safety significance (Green) and a Non-Cited Violation of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed on April 16, 2018, for the licensees failure to establish a preventive maintenance procedure for the safety-related degraded voltage relays that was appropriate to the circumstances. Specifically, the licensee failed to ensure that the first-time functional test and calibration for relay 1274B241 (Procedure MAQC773524, Quad Cities NOAD Unit 2 Tech Spec Undervoltage Relay and Degraded Voltage Relay Calibration) was at an appropriate frequency to ensure that the relay would perform its Technical Specification function.
05000254/FIN-2018001-012018Q1Quad CitiesRepeat Use of Written Exams During Licensed Operator Requalification ExaminationsThe inspectors identified a Severity Level IV Non-Cited Violation of 10 CFR 55.49, Integrity of Examinations and Tests, due to the licensee engaging in an activity that compromised the integrity of an examination. Specifically, the Quad Cities 2015 Licensed Operator Requalification (LOR) written examinations were duplicated from the 2013 LOR written examinations, the 2017 LOR written examinations were duplicated from the 2015 LOR examinations, and four individuals were administered the same written examinations from the previous requalification examination cycle.
05000265/FIN-2018001-022018Q1Quad CitiesFailure to Establish Design Standard for Unit 2 Residual Heat Removal Service Water PumpsThe inspectors identified a finding of very low safety significance (Green) and a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that the design bases standard and other requirements necessary to assure adequate quality were included in the design documents for the Unit 2 residual heat removal service water pumps. Consequently, the licensee failed to ensure the Unit 2 pumps were designed and constructed in accordance with the Standards of the Hydraulic Institute as identified in the Updated Final Safety Analysis Report.
05000254/FIN-2018001-032018Q1Quad CitiesHalf Scram Due to Low Voltage on 24/48 Vdc SystemA finding of very low safety significance (Green) and a Non-Cited Violation of Technical Specification 5.4.1, Procedures, was self-revealed on January 11, 2018, for the licensees failure to perform an equalizing charge on the Unit 1B 24/48 Vdc battery prior to returning the 24/48 Vdc battery to a normal configuration following a test discharge, which was required by station procedures. The failure to follow procedures led to a low voltage condition and caused a Unit 1B channel half scram in the reactor protection system.
05000440/FIN-2018001-012018Q1PerryFailure to Notify the NRC within 60 Days of a Condition Prohibited by Technical SpecificationsThe inspectors identified a Severity Level IV Non-Cited Violation of 10 CFR 50.73, Licensee Event Report System, for the licensees failure to report a condition that was prohibited by the plants Technical Specifications to the U.S. Nuclear Regulatory Commission (NRC) within 60 days. Specifically, the licensee did not report a condition that, as determined by the NRC, rendered the Division 2 Diesel Generator (DG) inoperable for a period longer than the Technical Specification allowed completion times of its associated required actions.
05000254/FIN-2018001-042018Q1Quad CitiesEnforcement Action: EA18021: EDG Non-conformance for Tornado Missiles (EGM 15002)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 EGM applies specifically to a structure, system, and component (SSC) that is determined to be inoperable for tornado-generated missile protection. 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 Quad Cities Nuclear Generating Station, the EGM provided for enforcement discretion of up to 3 years from the original date of issuance of the EGM. The EGM allowed NRC staff to exercise this enforcement discretion only when a licensee implements, prior to the expiration of the time mandated by the limiting conditions for operation (LCO), initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. In addition, licensees were expected to follow these initial compensatory measures with more comprehensive compensatory measures within approximately 60 days of issue discovery. The comprehensive measures should remain 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. In 1967, the NRC issued general design criterion to which the Quad Cities Nuclear Generating Station was evaluated against. Quad Cities Updated Final Safety Analysis Report (UFSAR), Section 3.1, Conformance with NRC General Design Criteria, discusses this criterion and its applicability to the sites design. Specifically, UFSAR Section 3.1.1.2, Criterion 2Performance Standards, states, those systems and components essential to the prevention of accidents or to mitigation of their consequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects. Section 3.1.1.2 further states that plant equipment which is important to safety is designed to permit safe plant operation and to accommodate all design basis accidents for all appropriate environmental phenomena at the site without loss of their capability. On March 1, 2018, during an engineering review of the Quad Cities, Units 1 and 2 facility design, the licensee identified a nonconforming condition with the aforementioned general design criterion. Specifically, the licensee identified that the three EDG systems intake stacks, exhaust stacks, fuel oil storage tank vent lines, and diesel oil day tank vent lines were inadequately protected against tornado missiles. As a result of the nonconforming condition, the licensee declared the Units 1, 2, and 12 EDG systems inoperable and entered the Technical Specifications (TS) LCO required action statements. The condition was reported to the NRC in Event Notice 53235 as an unanalyzed condition and a condition that could have prevented fulfillment of a safety function. Corrective Actions: The licensee documented the inoperability and functionality of the affected SSCs and the applicable TS LCO action statements 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 as operable but nonconforming prior to expiration of the required LCO action statements. The licensees initial (and final) compensatory measures included: verification that procedures and training for a tornado watch or warning were in place to provide additional instructions for operators to respond in the event of tornados 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 emergency response procedures; establishment of a heightened level of station awareness and preparedness relative to identifying tornado missile vulnerabilities; and revision to procedure QCOA 001010, Tornado Watch-Warning, Severe Thunderstorm Warning, or Severe Winds, to include guidance for unobstructing and/or repairing crimped diesel fuel oil tank vent lines. Corrective Action References: IR 1281009: Tornado Missile Protection Unresolved Item and IR 4110003: EDG Non-Conformance for Tornado Missiles Enforcement: Violation: 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); and TS 3.8.1: AC SourcesOperating. Severity/Significance: The subject of this enforcement discretion, associated with tornado missile protection deficiencies, was determined to be less than red (i.e., high safety significance) based on a generic and bounding risk evaluation performed by the NRC in support of the resolution of tornado-generated missile non-compliances. The bounding risk evaluation is discussed in Enforcement Guidance Memorandum 15002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Non-Compliance, and can be found in ADAMS Accession No. ML16355A286. Basis for Discretion: The NRC exercised enforcement discretion in accordance with Section 2.3.9 of the Enforcement Policy and EGM 15002 because the licensee initiated initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. The licensee reviewed their initial compensatory measures to determine if more comprehensive compensatory measures were warranted. Upon their review, the licensee concluded that their initial compensatory measures were sufficient to satisfy both the short-term and long-term actions required by the EGM and therefore no additional actions were necessary for enforcement discretion. The disposition of this enforcement discretion closes URI05000254/201100904; 05000265/ 201100904: Tornado Missile Protection of the Emergency Diesel Generator Air Intake and Exhaust.
05000315/FIN-2017004-042017Q4CookFailure to Verify the Adequacy of the Design for a Temporary ModificationA finding and associated violation of 10 CFR 50 Appendix B Criterion III self-revealed when licensee personnel could not obtain a water sample from a location designated as a connection point for a safety related temporary modification. Specifically, the licensee developed a temporary modification to add water to CCW but failed to verify the adequacy of the design in that the licensee did not validate the connection point could supply sufficient water as a source for CCW make-up. As an immediate action the licensee reestablished flow through the valves. The inspectors determined that the licensees failure to verify the adequacy of the design for the temporary modification was more than minor because it was associated with equipment performance attribute of Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance (Green) because the finding affected the qualification of CCW but did not render it inoperable. In this case, CCW remained operable based on credit taken for isolation valve capability. The finding includes a cross-cutting aspect in the human performance area of H.14, Conservative Bias.
05000315/FIN-2017004-032017Q4CookFailure to Promptly Correct The CAQ by Not Testing the CCW Leak Isolation ValvesThe inspectors identified a finding and associated NCV of Title 10 of the Code of Federal Regulations (CFR) Title 50, Appendix B Criterion XVI for failing to promptly correct a condition adverse to quality (CAQ). Specifically, in Inspection Report (IR) 05000315/3162015008 the NRC issued an NCV of 10 CFR 50 Appendix B Criterion III for the licensees failure to leak test isolation valves between redundant trains of the component cooling water (CCW) systems for Units 1 and 2. Despite opportunities to restore compliance, for Unit 1, the licensee suffered the violation from November 17, 2015, through November 4, 2017. As of December 31, 2017, the licensee continues to be in violation on Unit 2. The licensee tested the Unit 1 isolation valves during the fall 2017 outage and has scheduled testing of the Unit 2 valves in the spring 2018 outage. The inspectors determined that the licensees failure to promptly correct the CAQ by not testing the CCW leak isolation valves or otherwise restoring compliance was more than minor. The inspectors determined the issue was more than minor because it adversely affected the Mitigating Systems cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The issue was not greater than green because it did not render CCW inoperable. The inspectors determined the finding included a cross-cutting aspect of H.1, Resources.
05000315/FIN-2017004-022017Q4CookUnit 1 Letdown System Safety Valve Lift During Preparations for CooldownRefueling Outage Activities a. Inspection Scope The inspectors reviewed the Outage Safety Plan (OSP) and contingency plans for the Unit 1 refueling outage (RFO), conducted September 13 through November 26, to confirm that the licensee had appropriately considered risk, industry experience, and previous site-specific problems in developing and implementing a plan that assured maintenance of defense-in-depth. During the RFO, the inspectors observed portions of the shutdown and cooldown processes and monitored licensee controls over the outage activities listed below: licensee configuration management, including maintenance of defense-in-depth commensurate with the OSP for key safety functions and compliance with the applicable TS when taking equipment out of service; implementation of clearance activities and confirmation that tags were properly hung and equipment appropriately configured to safely support the work or testing; installation and configuration of reactor coolant pressure, level, and temperature instruments to provide accurate indication, accounting for instrument error; controls over the status and configuration of electrical systems to ensure that TS and OSP requirements were met, and controls over switchyard activities; monitoring of decay heat removal processes, systems, and components; controls to ensure that outage work was not impacting the ability of the operators to operate the spent fuel pool cooling system; reactor water inventory controls including flow paths, configurations, and alternative means for inventory addition, and controls to prevent inventory loss; controls over activities that could affect reactivity; maintenance of secondary containment as required by TS; licensee fatigue management, as required by 10 CFR 26, Subpart I; refueling activities, including fuel handling and reactor assembly/disassembly; startup and ascension to full power operation, tracking of startup prerequisites, walkdown of the containment to verify that debris had not been left which could block emergency core cooling system suction strainers, and reactor physics testing; and licensee identification and resolution of problems related to RFO activities. Documents reviewed are listed in the Attachment to this report. Inspections activities performed in the third quarter coupled with those in the fourth quarter constituted one RFO sample as defined in IP 71111.2005. b. Findings (Opened) Unresolved Item 05000315/201700402, Unit 1 Letdown System Safety Valve Lift During Preparations for Cooldown Introduction: Shortly after the shutdown for the Unit 1 refueling outage in September 2017, the licensee was establishing conditions in the charging and letdown system for the upcoming cooldown. After lowering letdown flow and attempting to adjust pressure, a letdown safety valve lifted and failed to completely reseat. Review of plant parameters following the event revealed that the evolution created saturation conditions in the letdown system. Subsequently, the steam bubbles collapsed causing a water hammer that lifted and damaged a relief in the system. The event was discussed in Section 4OA3 of Inspection Report 05000315/05000316/2017003. Description: The inspectors reviewed the licensees follow up of the issue in the CAP and spoke to personnel in the operations and maintenance departments. The licensee identified potential issues in the areas of procedure adequacy, operator performance, and equipment performance. However, the inspectors could not reconcile information on plant conditions with licensees statements regarding the cause. Because of ambiguity regarding the cause, the inspectors could not determine whether the corrective actions taken by the licensee were adequate. The licensee determined that an apparent cause evaluation need not be done therefore the inspectors reviewed available data, including plant computer data and a prior event from 2004. Since it is unclear what, if any, performance deficiency exists associated with this issue, the inspectors determined an unresolved item (URI) was necessary pending further follow up of the issue.Following the lifting of the safety valve, the licensee isolated letdown to stop the remaining leakage through the valve. The licensee then cycled the valve sufficiently enough for it to reseat so letdown could be restored and the cooldown continued. The safety valve was later discovered to be damaged from the event, so it was also repaired. Walkdowns were also conducted of the letdown piping to ensure no damage had occurred during the pressure transient. As part of their corrective actions, the licensee made some changes to the letdown procedure, recalibrated a letdown flow control valve, and developed actions to cover the event and lessons-learned in training. However, as stated above, the inspectors were unable to determine if these were sufficient to address the prevailing cause of the issue. The inspectors developed a series of questions for the licensee to explore more of the details behind the various potential issues. In order close the URI, the inspectors need to review the licensees response to questions provided and review available documentation of the event. (URI 05000315/201700402, Unit 1 Letdown System Safety Valve Lift During Preparations for Cooldown)
05000315/FIN-2017004-012017Q4CookFailure to Correct Numerous Anchor Darling Double Disc Gate Valve Non-ConformancesThe inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action for the licensees failure to correct a design non-conformance reported to the licensee through two related 10 CFR Part 21 reports. In March 2013, the licensee identified that 28 safety-related Anchor Darling double disc gate valves (ADDDGVs) may not have been assembled with an assumed amount of valve stem to wedge pre-torque before the stem was pinned into the wedge. The licensee had restored compliance to only one of these valves and had no plans to restore quality to the remaining 27 valves prior to the inspection. The licensee entered the inspectors conclusions into their corrective action program (CAP) as AR 201710399. At the end of this inspection the licensees plan was to restore compliance by either correcting the Part 21 issue or changing the design to accept the stem not having any pre-torque into the wedge.The performance deficiency was determined to be more than minor because if left uncorrected could become a more significant safety concern. Specifically, the failure to correct the design deficiencies could result in the valve pin breaking and consequential valve damage if the valves were operated at a high enough torque and/or thrust value(s). 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 an operability determination which concluded that all 28 valve wedge pins had not sheared based upon the known historic operational history, pin material properties, and for using stem to wedge thread friction in some cases. The inspectors determined that this finding was not indicative of recent performance and therefore did not have a cross-cutting aspect assigned.
05000315/FIN-2017004-052017Q4CookLicensee-Identified ViolationThe following violation of very low significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as an NCV. Title 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings states, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings appropriate to the circumstances and shall be accomplished in accordance with those instructions, procedures, and drawings. Equipment tagging is a safety related process implemented by procedure 12OHP2110CPS001, Clearance Permit System. Contrary to 12OHP2110CPS001 step 4.4.3, which directs operators to comply with the tagout on the Unit 2 East Motor Driven AFW pump room cooler, the operators mistakenly secured and tagged the Unit 1 East Motor Driven AFW pump room cooler instead. This rendered the Unit 1 East Motor Driven AFW pump inoperable. The violation occurred at 0219 on September 6, 2017, and concluded at 0623 the same day after the error was realized and corrected. The licensee entered the issue into their CAP as AR20178509. The finding screened to Green because there was no loss of system function, nor loss of a train for greater than the Technical Specification allowed outage time.
05000461/FIN-2017003-052017Q3ClintonFailure to Establish Secondary Containment Prior to Entering MODE 2The inspectors documented a self-revealed finding of very low safety significance and an associated NCV of TS LCO 3.0.4, for the failure to follow station procedure CCAA201, Plant Barrier Control Program, Revision 11. Specifically, the licensee entered MODE 2 from MODE 4 without meeting the requirements of LCO 3.0.4 for entering a mode when an applicable LCO is not met. The licensee had not met LCO 3.6.4.1 because the doors to the B reactor water cleanup room were both opened instead of being closed to make secondary containment operable as required in MODE 2. The licensee entered this issue into their CAP as AR 04017613. As corrective actions, the licensee planned to conduct training for site personnel.The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it impacted the Barrier Integrity cornerstone attribute of configuration control and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to follow the station procedure by not identifying that the open doors required a plant barrier impairment (PBI) permit that would have identified the doors as a constraint to entering MODE 2 resulted in the unit transitioning to MODE 2 with the secondary containment inoperable. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, October 7, 2016, the finding was screened against the Barrier Integrity cornerstone and determined 5 to be of very low safety significance because the finding only represented a degradation of a radiological barrier function provided for auxiliary building. The inspectors determined that this finding affected the cross-cutting are of human performance in the aspect of training, where the organization provides training and ensures knowledge transfer to maintain a knowledgeable, technically competent work force and instill nuclear safety values. Specifically, station personnel did not know the process for routing a PBI permit and did not know when a PBI permit was required. (H.9)
05000461/FIN-2017003-042017Q3ClintonFlow Control Valves Not Locked Out Results in Reactor Recirculation Pump RunbackThe inspectors documented a self-revealed finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1, Procedures, for the licensees failure to establish sufficient instructions in station procedure Clinton Power Station (CPS) 3103.01, Feedwater (FW), Revision 31e, for changing modes of operation for the nuclear steam supply system. Specifically, the station procedure did not provide instructions requiring the locking out the flow control valves (FCVs) to prevent a reactor recirculation FCV runback while changing the feedwater pump lineup resulting in an unexpected plant transient and 9.2 percent change in reactor power. The licensee entered this issue into their corrective action program (CAP) as Action Request (AR) 04007861. As corrective actions, the licensee revised their CPS 3103.01 procedure to require that the FCVs be locked out prior to shifting reactor feed water pumps. The performance deficiency was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012,because the finding 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 and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to have adequate procedures for shifting feedwater pumps during a plant shutdown on May 7, 2017, resulted in an unexpected recirculation pump run back and a 9.2 percent change in reactor power. Using IMC 0609, Attachment 4, Initial Characterization of Findings, andAppendix A, The Significance Determination Process for Findings At-Power, issuedJune 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because the event did not cause a reactor scram. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of conservative bias, where individuals use decision making practices that emphasize prudent choices over those that are simply allowable and a proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the procedure provided for the option to lockout the reactor 3 recirculation flow control valves if deemed necessary during a shift of the reactor feedwater pumps and the operations crew did not make the prudent choice of locking out the valves before determining that it was safe to proceed. (H.14)
05000461/FIN-2017003-032017Q3ClintonFailure to Perform Engineering Evaluation to Determine the Cause of Failure of SnubbersThe inspectors identified a finding of very low safety significance and an associated NCV of Title 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to demonstrate compliance with the requirement as prescribed in procedure ERCL330, CPS Snubber Program, Revisions 1 and 2. Specifically, the licensee failed to perform engineering evaluations to determine the cause of failure of snubbers that did not satisfy their functional testing acceptance criteria. The licensee entered this issue into their CAP as ARs 04015242 and 04041302. As corrective actions, the licensee evaluated the components affected by the failed snubber and determined that no operability issues existed. The performance deficiency was determined to be more-than-minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it was associated with the Mitigating Systems cornerstone attribute of Protection against External Factors and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability for mitigating systems to respond to initiating events. Specifically, compliance with ERCL330 would ensure the failed snubber wasevaluated for the cause of failure, to ensure the licensee identified other snubbers that may have been vulnerable to the same type of deficiency. This would ensure that any potential undesired loading on the piping system could be avoided and the affected safety-related residual heat removal and reactor water cleanup piping systems could continue to perform their design function of maintaining the pressure boundary and structural integrity following a postulated design basis seismic event. The inspectors determined the finding could be evaluated using the Significance Determination Processin accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, for the Mitigating Systems cornerstone and then Exhibit 4, External Events Screening Question. The finding screened as having very low safety significance because in each instance, the inspectors answered No to Questions 1 and 2 ofExhibit 4. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of consistent process, where individuals use a consistent, systematic approach to make decisions. Specifically, the licensee failed to establish a systematic approach to evaluating snubbers that did not meet the acceptance criteria to ensure all required aspects were addressed. (H.13)
05000461/FIN-2017003-022017Q3ClintonFailure to Adequately Control Access in Locked High Radiation AreaA finding of very low safety significance and an associated NCV of TS 5.4.1 was self-revealed when individuals failed to adequately control access in locked high radiation areas (LHRAs). Specifically, the failure to meet all of the requirements of Procedure RPAA460, Attachment 5, represented a failure to comply with Radiation Work Permit CL1700518, C1R17 (Drywell) DW Bioshield Inservice Inspection Activities. This resulted in four individuals entering a LHRA that they had not been specifically authorized to enter. These individuals entered the incorrect location and were inside the area for approximately 2-3 minutes before they noticed that they were in the incorrect area. The individuals knew that they were in the incorrect location when they could not find the nozzles that they planned on inspecting. The individuals exited the area and were simultaneously told to exit the area by the radiation protection technician (RPT) providing remote coverage which demonstrated that the four workers were not in the authorized work area. Immediate corrective actions taken by the licensee included immediately suspending the work that was scheduled to take place within the bioshield associated with this job. Electronic dosimeters and dosimeters were immediately collected from the individuals that entered the area so the dose that was received could be known. The licensee also interviewed all the individuals that were involved in this bioshield entry, and the RPT that performed the brief. These interviews were conducted to understand which parts of the process associated with entry into LHRAs failed and led to this event transpiring. The licensee entered this event into their CAP as AR 04012075. As corrective actions the licensee planned to observe high radiation area and locked high radiation area briefs, for both in house and traveling RPTs. The licensee also planned to modify the bioshield as-low-as-reasonably-achievable (ALARA) plan template to label all accessible bioshield doors with elevation and azimuth.The inspectors determined that the performance deficiency was more-than-minor in accordance with IMC 0612, Appendix B, 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, in that, the workers entered an area that required the radiation dosimeter to be relocated to the workers knee, and the workers were wearing them on the head for the intended work location. The finding was determined to be of very-low safety significance (Green) in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, because: (1) it did not involve as-low-as-reasonably-achievable 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 this finding affected the cross-cutting area of human performance in the aspect of resources, where leaders ensure that personnel, 6 equipment, procedures and other resources are available and adequate to support nuclear safety. Specifically, radiation protection leadership failed to ensure that the RPT was capable of meeting the expectations for performing the LHRA briefing in accordance with station procedure RPAA460, Attachment 5. (H.1)
05000461/FIN-2017003-012017Q3ClintonMSIV TS Leakage Limits Exceeded Due to Condition Based Maintenance ApproachThe inspectors documented a self-revealed finding of very low safety significance and an associated NCV of TS limiting condition for operation (LCO) 3.6.1.3, for the failure to follow station procedure ERAA200, Preventative Maintenance Program, Revision 3. Specifically, the licensee utilized a condition-based maintenance approach on the main steam isolation valves (MSIVs) that failed to monitor and trend equipment performance so that planned maintenance could be performed prior to the MSIVs exceeding the TS leakage limits. The licensee entered this issue into their CAP as AR 04009845. As corrective actions, the licensee repaired and tested the valves prior to returning the unit to the modes of applicability.The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it impacted the Barrier Integrity cornerstone attribute of configuration control and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the monitoring and trending of local leak rate tests on the MSIVs did not provide performance data that would allow planned maintenance to the valves prior to the valves failing resulting in exceeding TSleakage requirements for the MSIVs. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, October 7, 2016, the finding was screened against the Barrier Integrity cornerstone Reactor Containment and did represent an actual open pathway in the physical integrity of reactor containment. The inspectors proceeded to Appendix H, Containment Integrity Significance Determination Process, and determined that it was a Type B finding that was related to a degraded condition that has potentially important implications for the integrity of the containment, without affecting the likelihood of core damage. The inspectors used Figure 6.1, Road Map for LERF based Risk Significance for Evaluation of Type-B Findings at Full Power and determined this finding is of very low safety significance (Green). The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of design margins, where the organization operates and maintains equipment within design margins. Special attention is placed on maintaining fission product barriers, defense-in-depth and safety related equipment. Specifically, the procedure for testing the MSIVs utilized an administrative limit that provided no margin to correct performance prior the valves becoming inoperable. (H.6)
05000373/FIN-2017001-032017Q1LaSalleLicensee-Identified ViolationTitle 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and drawings, requires, in part, that activities affecting quality be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to this, on February 6, 2017, the licensee failed to accomplish an activity affecting quality in accordance with licensee procedure, CC AA 201, Revision 11, Plant Barrier Control Program. Specifically, the licensee failed to implement compensatory actions required by the Plant Barrier Control Program which resulted in multiple doors being impaired at the same time such that safety -related equipment in the Unit 2 Division II switchgear room and Unit 2 749 Auxiliary Building were declared inoperable. The licensee documented the issue in their CAP as Action Request (AR ) 3972830. The inspectors determined that this issue was of very low safety significance because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component ( SSC ); (2) did not represent a loss of system and/or function; (3) did not represent the actual loss of safety 35 function of at least a single train for greater than its technical specification ( TS ) allowed outage time; (4) did not represent an actual loss of one or more non TS trains of equipment during shutdown designated as risk significant for greater than 24 hours; and (5) did not degrade a functional auto- isolation of residual heat removal ( RHR) .
05000373/FIN-2017001-012017Q1LaSallenadequate Controls for ASME Code VT 3 Internal Examination of Pumps and ValvesGreen . The inspectors identified a finding of very-low safety significance with an associated NCV of Title 10 of the Code of Federal Regulations (CFR) , Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because the licensee failed to establish a procedure that ensured the American Society of Mechanical Engineers (ASME) Code VT 3 examination of the internal surface of valves or pumps occurred in the as -found condition (e.g., prior to repairs). Consequently, the licensee repaired internal damage to the 2B33 F067B valve prior to the Code VT 3 examination which potentially resulted in an ineffective VT 3 examination. The licensee entered this issue into their corrective action program (CAP) as Action Request (AR) 3972620, initiated actions to complete another VT 3 examination of valve 2B33 F067A or valve 2B33 F067B during the current outage and was evaluating additional controls for scheduling VT 3 internal examinations of pumps and valves. The performance deficiency was determined to be more- than- minor because it affected the Initiating Events cornerstone attribute of equipment performance and adversely affected the cornerstone objective to l imit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, if left uncorrected, this finding would lead to a more significant safety concern because it increa sed the likelihood of an operational challenge to the plant caused by a recirculation system line break initiated from undetected service -induced defects left in service inside pumps or valves as a result of ineffective VT 3 examinations. The finding was screened in accordance with Inspection Manual Chapter 0609, Appendix A, and the inspectors answered No to the applicable Phase 1 Initiating Events Screening question because the finding did not result in a reactor trip and/or loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Therefore, this finding was determined to have very- low safety significance (Green) . The finding had a cross -cutting aspect of Work Management in the Human Performance cross -cutting area because licensee managers failed to establish an adequate process of planning, controlling, and executing 3 work activities such that nuclear safety is the overriding priority as evidenced by the lack of appropriately controls for scheduling the VT 3 internal examination of the 2B33 F067B valve (H.5)
05000373/FIN-2017001-022017Q1LaSalleFailure to Perform Preventive Maintenance Resulted in Stem -to-Disc Separation of Safety -Related ValveGreen . A finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self -revealed for the licensees failure to ensure that activities affecting quality were prescribed in a manner appropriate to the circumstances for the Unit 2, Division 3 , diesel generator (DG) system . Specifically, the licensees processes for the control and administration of preventive maintenance (ER AA 200/WC AA 120) failed to ensure that safety -related valve, 2E22 F319, the 2B DG cooling water strainer backwash valve, was replaced or refurbished at a frequency that would prevent corrosion- related stem -to-disc separation. The licensee entered this issue int o the ir CAP as AR 1122320. Corrective actions planned and completed included replacement of the 2E22 F319 valve with a stainless steel design and performing an apparent cause evaluation of the degraded condition. The performance deficiency was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage) . Specifically, the failure to perform preventive maintenance on the 2E22 F319 valve resulted in a degraded condition which adversely affected the reliability of the high pressure core spray system to respond to an initiating event. The inspectors evaluated the finding using the significance determination process in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, dated June 19, 2012. The inspectors reviewed the Mitigating Systems screening questions in Exhibit 2 and answered No to question A.1, If the finding is deficiency affecting the design or qualification of a mitigating SSC (structure, system, or component) , does the SSC maintain its operability or functionality. The inspectors answered Yes to question A.2, Does the finding represent a loss of system and/or function; therefore, a detailed risk evaluation was required. The detailed risk evaluation determined that the finding screened as having very low safety significance (Green). This finding had a cross -cutting aspect in the area of Problem Identification and Resolution, because t he organization failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance (P .3).
05000254/FIN-2017001-012017Q1Quad CitiesFailure to Ensure Hardware Secure for Breaker MOC Switch LinkageGreen . A finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion V was self -revealed on January 27, 2017, when the Unit 1C residual heat removal service water (RHRSW ) pump was started for a routine surveillance evolution and all expected annunciators and equipment failed to operate properly, which led to the licensee declaring the Unit 1C RHRSW pump inoperable. Specifically, t he licensee failed to establish a procedure for the mechanism operated contact (MOC) switch linkage arm that was appropriate to the circumstances to ensure the component would c ontinue to perform its function. Immediate corrective actions included reconnecting the MOC switch linkage arm assembly and test ing it by starting the 1C RHRSW pump prior to declaring the pump operable. In addition, the licensee planned procedure revisions to QCEPM 0200 11 that would specify a torque value to ensure the MOC switch linkage arm was adequately secured and could perform its function. Th is issue was entered into the licensees corrective action program as Issue Report 3967424 . The finding was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure the MOC switch linkage arm was adequately fastened led to the failure of the component and its associated Unit 1C RHRSW pump d uring breaker operation on January 27, 2017. T he finding was determined to be of very low safety significance (Green), because the inspectors answered No to all of the questions in IMC 0609, Appendix A, The Significance Determination Process for Findings at Power , Exhibit 2, Mitigating Systems Screening Questions, Section A, Mitigating SSCs and Functionality. The inspectors determined this finding affected the cross- cutting area of human performance, in the aspect of avoid complacency, which state s, Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee failed to recognize a potential risk and inherent latent issue for a condition identified in 2015 at Quad Cities, when a MOC switch failed to perform its function due to a missing nut in a different breakers linkage assembly. The licensee identified and corrected the 3 condition but failed to evaluate the cause of the missing nut because it did not impact the operability of the component . I n the 2015 instance, the MOC switch issue only affected indications for the component and had no adverse impact on the ability of the component to perform its function (H.12 ).
05000237/FIN-2016004-022016Q4DresdenLicensee-Identified ViolationTitle 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2) requires that a holder of a nuclear power reactor operating license follow and maintain the effectiveness of an emergency plan that meets the requirements in 10 CFR Part 50, Appendix E and the planning standards of 10 CFR 50.47(b). Title 10 CFR Part 50.47(b)(4) states, A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures. Contrary to the above, between April 2013, and February 2016, the licensee failed to maintain the effectiveness of the emergency plan by failing to maintain the effluent parameters contained in the standard emergency classification and action level scheme. Specifically, the standard emergency classification and action level scheme associated with the radiological effluents at Dresden Nuclear Power Station was not updated to reflect the changes in the X/Q dispersion factor that were made during the April 2013, Offsite Dose Calculation Manual revision. Consequently, the effluent monitor emergency classification and action level thresholds were non-conservative by a factor of 3.8 until this condition was identified and corrected by Dresden Nuclear Power Station in February 2016. The inspectors determined that the finding was of very low significance (Green) in accordance with NRC Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, Figure 5.41, because the emergency action level classification of an Unusual Event, RU1, would be declared in a degraded manner, not within the required 15 minutes. The emergency action level classification for the Alert, Site Area Emergency, and General Emergency (RA1, RS1, and RG1) would still be capable of being declared in timely manner, within 15 minutes, using alternate conditions within the emergency action level. Because this finding is of very low safety significance, and has been entered into Exelons CAP under IR 02652711, this violation is being treated as a Green NCV consistent with Section 2.3.2 of the NRCs Enforcement Policy.
05000237/FIN-2016004-012016Q4DresdenFailure to Comply With Radiation Work Permit Requirements Resulting In Unplanned Dose Rate AlarmsA finding of very-low safety significance, and an associated Non-Cited Violation (NCV) of Technical Specification 5.4.1 was self-revealed when workers violated a radiation work permit (RWP) by entering an area that was outside of the scope of the original RWP brief without obtaining a required appropriate brief, resulting in these workers receiving unplanned electronic dosimeter dose rate alarms. These workers immediately exited the area and reported the event to the radiation protection staff. The licensee entered these issues as two separate events into their CAP as Issue Reports (IR) 02735594 and IR 02735651. The inspectors determined that the performance deficiency was more than minor in accordance with Inspection Manual Chapter 0612, Appendix B, 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, worker entry into areas beyond the RWP briefing could lead to unintended dose. The finding was determined to be of very-low safety significance (Green) in accordance with Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, because: (1) it did not involve as-low-as-reasonably-achievable 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 concluded that the cause of the finding involved a cross-cutting component in the human performance area of challenging the unknown because the individual did not stop when faced with an uncertain condition. Risks were not evaluated and managed before proceeding (H.11).
05000456/FIN-2016004-012016Q4BraidwoodInadequate Control of Welding During FW System Pipe ReplacementA finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion IX, Control of Special Processes, was identified by the inspectors for the licensees failure to assure that thermo couple (TC) attachment welding was controlled and accomplished by qualified personnel using qualified procedures and to assure that the post-TC attachment weld removal non-destructive examination (NDE) was incorporated into Work Order (WO) 01836557 that provided instructions to replace a pipe segment in the safety-related portion of the feedwater (FW) system. The licensee corrective actions for this finding included documenting this issue as a potential violation of NRC requirements in Issue Report (IR) 02728742, removal of the unqualified welds, and issuing revisions to WO 01836557 that included licensee-approved weld procedures and surface examinations of FW pipe affected by unqualified TC welds. This finding was determined to be of more than minor significance because it affected the Reactor Safety Initiating Events Cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In particular, if left uncorrected this issue would have the potential to lead to a more significant safety concern because it increased the likelihood of an operational challenge to the plant caused by a FW system line break induced by cracking initiated from unqualified welds. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At Power, Exhibit 1, Initiating Events Screening Questions. Under Part B, Transient Initiators, of the Exhibit 1 questions, the inspectors answered No because the finding did not result in a reactor trip and/or loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Therefore, this finding was screened as having very low safety significance (Green). This finding had a cross-cutting aspect of Field Presence in the cross-cutting area of Human Performance since licensee managers failed to provide adequate oversight of site and vendor personnel to assure that the TC attachment welding was controlled and accomplished by qualified personnel using qualified procedures and to assure that the post-TC attachment weld removal NDE was incorporated into WO 01836557. (H.2)
05000237/FIN-2016009-012016Q3DresdenMain Steam Acoustic Safety/Relief Valve Monitoring Channel Calibration Not PerformedThe inspectors identified a finding of very-low safety significance for the failure to perform a 24-month channel calibration of the Regulatory Guide 1.97 safety/relief valve acoustic monitoring system in accordance with the Technical Requirements Manual. Specifically, the licensee failed to perform a channel calibration, where the channel calibration shall encompass all devices in the channel required for channel operability and the channel functional test. The performance deficiency was determined to be more-than-minor because the finding was associated with the Mitigating Systems cornerstone attribute of Procedure Quality and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to maintain the acoustic safety/relief valve position indicators instrumentation in accordance with the Technical Requirements Manual. The performance deficiency affected the design or qualification of a mitigating system, structure or component; however, the system, structure or component maintained its functionality based on successful completion of channel functionality checks. Since the system, structure or component remained functional, the inspectors screened the finding as having very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of the licensees current performance.
05000440/FIN-2016007-032016Q2PerryFailure to Comply With ASME Code Requirements for Repair on Code Class 1 ComponentA finding of very-low safety significance (Green) and associated NCV of 10 CFR 50.55a(g)(4) was identified by the inspectors for the licensees failure to maintain the American Society of Mechanical Engineers (ASME) Code Class 1 component in accordance with ASME Code Section XI requirements. Specifically, the licensee failed to measure and document the method of measuring the cavity created after removal of indications on the reactor water clean-up line prior to return to service. The inspectors determined that the licensees failure to maintain the ASME Code Class 1 component in accordance with ASME Code Section XI requirements was a performance deficiency. This performance deficiency was found to be more-than-minor, and a finding, because the performance deficiency, if left uncorrected could become a more significant safety concern. Specifically, absent NRC identification, the licensee would not have questioned the potential challenge to component functionality since the cavity measurements were not performed. This condition could potentially lead to the failure of the reactor water clean-up bottom head drain, which in turn, could lead to a potential loss of reactor coolant. The inspectors reviewed the finding using Attachment 0609.04, Initial Characterization of Findings, Table 3 SDP Appendix Router. The inspectors answered No to the question in Section A of Table 3 and therefore the finding was evaluated using the SDP in accordance with IMC 0609, The Significance Determination Process (SDP) for At-Power Operations, Appendix A, Exhibit 1, Initiating Events Screening Questions. The inspectors answered No to the questions in Exhibit 1 and determined this finding to have a very-low safety significance (Green). The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, Design Margin, for the licensees failure to maintain equipment within design margins. Specifically, the licensee staff failed to ensure that metal removal performed on an ASME Code Class 1 component did not result in a condition where the minimum design wall thickness of the component was compromised, and therefore, failed to ensure design margin was maintained.
05000440/FIN-2016007-022016Q2PerryUse of Unapproved Standard for Site Flooding Modifications and AnalysisThe inspectors identified a Severity Level IV, NCV of 10 CFR 50.59, Changes, Tests, and Experiments, having very-low safety significance (Green) for the licensees failure to conclude that site flooding modifications and associated analysis included a standard that resulted in a departure from the method of evaluation as described in the Updated Final Safety Analysis Report. Specifically, the licensee used a new method for evaluation of design basis flooding at Perry Nuclear Power Plant that is different from the method described in the Updated Final Safety Analysis Report and not approved by the NRC. The inspectors determined that the licensees use of an unapproved methodology for site flooding modifications and associated analysis that constituted a departure from a method of evaluation was contrary to 10 CFR 50.59(c)(2)(8) and was a performance deficiency. Specifically, the licensee used a new method for evaluation of design basis flooding at Perry Nuclear Power Plant that is different from the method described in the Updated Final Safety Analysis Report and not approved by the NRC. The performance deficiency was determined to be more-than-minor, and a finding, because it affected the cornerstone attribute of protection against external factors and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In addition, the associated violation was determined to be more-than-minor because the inspectors determined that there was a reasonable likelihood that the changes would have required prior NRC approval. The inspectors determined that finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process. Using Attachment 0609.04, Initial Characterization of Findings, Table 2 the inspectors determined that the finding affected the Mitigating Systems cornerstone. As a result, the inspectors evaluated the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, for the Mitigating Systems cornerstone. The inspectors answered Yes to question A.1 in Exhibit 2 Mitigating Systems Screening Questions. Specifically, the inspectors determined the finding did not result in systems, structures, and components not being able to maintain their operability or functionality. Therefore, the inspectors determined the significance of this finding to be of very-low safety significance (Green). In accordance with Section 6.1.d of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very-low safety significance (i.e., green finding). The inspectors determined that this finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Problem Identification, for the licensees failure to identify issues completely, accurately, and in a timely manner. Specifically, the licensees 50.59 review committee failed to accurately identify the methodology change concern in Evaluation 14-01234 during a review documented in CR2015-14025.
05000440/FIN-2016007-012016Q2PerryFailure to Document 50.59 Evaluation for Replacement of a Manual Action with an Automatic ActionThe inspectors identified a Severity Level IV, NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50.59, Changes, Tests, and Experiments, having very-low safety significance (Green) for failure to document the basis for performing a plant modification where a manual operator action was replaced with an automatic action. Specifically, the licensee did not evaluate whether adding a safety-related function to a nonsafety-related component was within the licensing basis of the facility. The inspectors determined that the failure to perform a 10 CFR 50.59 evaluation for Plant Modification 11-0794 was contrary to 10 CFR 50.59(d)(1) and was a performance deficiency. The performance deficiency was determined to be more-than-minor and a finding, because the finding impacted mitigating systems cornerstone attribute of Design Control and adversely affected the Cornerstone Objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, this plan modification added a Safety-Related function to a Nonsafety-Related component and, therefore, impacted the availability, reliability, and capability of the Safety-Related Battery Room ventilation system and the Safety-Related Motor Control Center, Switchgear, and Miscellaneous Electrical Equipment Area ventilation system. In addition, the associated violation was determined to be more-than-minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The inspectors determined that finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process. Using Attachment 0609.04, Initial Characterization of Findings, Table 2 the inspectors determined that the finding affected the Mitigating Systems cornerstone. As a result, the inspectors evaluated the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, for the Mitigating Systems cornerstone. The inspectors answered No to question A.4 in Exhibit 2 Mitigating System Screening Questions. Specifically, the inspectors determined the finding did not represent an actual loss of the Battery Room ventilation system or Motor Control Center, Switchgear, and Miscellaneous Electrical Equipment Area ventilation system because the automatic action had not been implemented at the time of the finding. Therefore, the inspectors determined the significance of this finding to be of very-low safety significance (Green). In accordance with Section 6.1.d of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very-low safety significance (i.e., green finding). The inspectors determined the finding was associated with the cross-cutting aspect of Procedure Adherence in the area of Human Performance, because the licensee failed to follow the screening criteria in Attachment 2 of Procedure NOBP-LP-4003A, FENOC 10 CFR 50.59 User Guidelines.
05000456/FIN-2016008-022016Q1BraidwoodFailure to Verify Air Intake for Diesel Driven Auxiliary Feedwater Pump was Adequately Protected from a High Energy Line BreakThe inspectors identified 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 III, Design Control, for the failure to verify the adequacy of the diesel driven Auxiliary Feedwater (AFW) pump design. Specifically, the licensee failed to verify the diesel driven AFW pump could perform its safe shutdown function following a high energy line break (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 its Corrective Action Program and took immediate corrective actions by declaring the diesel driven AFW pump inoperable and then implementing a temporary plant modification to relocate the diesel air intake to the Auxiliary Building where it is not susceptible to a HELB to restore operability of the pump. The licensees planned corrective actions are to complete a permanent plant modification to relocate the air intake to a location that is 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 associated 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 safety 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 3.4E-7/yr., which represents a finding of very-low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not indicative of the licensees current performance.
05000255/FIN-2015004-052015Q4PalisadesLicensee-Identified ViolationTitle 10 CFR 50.65(a)(1), requires, in part, that the holders of an operating license shall monitor the performance or condition of structures, systems, and components (SSCs), against licensee-established goals, in a manner sufficient to provide reasonable assurance that these SSCs, as defined in 10 CFR 50.65(b), are capable of fulfilling their intended functions. Title10 CFR 50.65(a)(2) states that monitoring as specified in 50.65(a)(1) is not required, where it has been demonstrated that the performance or condition of a SSC is being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remains capable of performing its intended function. Contrary to the above, as identified after the November 14, 2014, TDAFW pump trip, the licensee failed to demonstrate the performance or condition of the safety-related auxiliary feedwater system steam traps had been effectively controlled through the performance of appropriate preventive maintenance. Specifically, some of the safety-related steam traps, one relief valve, and one check valve associated with the steam supply piping of the turbine-driven AFW system were inappropriately classified in the maintenance rule program, resulting in inadequate and/or untimely maintenance being performed on these components, which probably contributed to the overspeed trip event. The licensee found 3 steam traps and one relief valve classified as non-critical components that were reclassified as high critical components and one steam trap and one check valve classified as run-to-failure components that were reclassified as high critical components. Some of these components also had no preventive maintenance (PM) strategies or ones that were not the correct frequency based on the component classification. The licensee identified this issue while conducting the equipment apparent cause evaluation for the overspeed trip event and documented actions to correct the issue in CR-PLP-2014-5477. The licensee performed inspections of all the steam traps required for the TDAFW pump operation and identified some issues with steam cutting, foreign material exclusion in the traps, and incomplete seat contact. These issues were corrected and PM changes have been made for all the system components mentioned above. The inspectors determined that the inconsistent equipment classifications and ineffective preventive maintenance strategy for the safety-related steam traps in the turbine-driven auxiliary feedwater system is considered a performance deficiency. The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Specifically, the licensee identified that the degraded condition of the moisture removal system could have led to excess condensate being present in the steam supply line which had the potential to adversely affect the operation of the turbine for the TDAFW pump, contributing to the overspeed trip event. The inspectors screened the issue using IMC 0609, Appendix A, The SDP for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, and answered Yes to the question of does this finding represent a loss of system and/or function? This trip of the TDAFW pump on overspeed was evaluated as a failure that impacted the ability of the AFW system to provide the specific function, which could only be accomplished by this train, of decay heat removal via steaming of the A Steam Generator. The turbine-driven AFW pump was also determined to not be in a condition to meet performance requirements defined by the probabilistic risk assessment success criteria, which for AFW is a 24 hour mission time. Therefore, the issue was screened further in a detailed risk evaluation. A Region III Senior Reactor Analyst performed a detailed risk evaluation using the NRCs Standardized Plant Analysis Risk Model for Palisades, Revision 8.20. The SRA assumed the turbine driven AFW pump was unavailable to perform its function for a period of 3 days because the pump was successfully tested and returned to service on November 16, 2014. Given the short exposure period, the calculated delta core delta frequency was less than 1.0E-7/yr. As a result of the low calculated delta core delta frequency, no additional analysis of external event risk contribution or large early release risk contribution was necessary. The dominant core damage sequence was a station blackout followed by the failure of the turbine driven AFW pump and the failure to recover onsite or offsite power. Therefore, the finding screened as very low safety significance (Green).
05000255/FIN-2015004-042015Q4PalisadesFailure to Perform a Required 50.59 Evaluation for Declassification of the CVCSThe inspectors identified a SL IV, NCV of 10 CFR, Part 50.59, Changes, Tests, and Experiments, and an associated finding of very-low safety significance (Green) for the licensees failure to maintain a record of the declassification of the Chemical Volume and Control System (CVCS) from safety-related to nonsafety-related, which includes a written evaluation that provides the bases for the determination that the change did not require a license amendment. The licensee entered this issue into their CAP, and after a review of the system, determined there was reasonable assurance that it could perform its function. The inspectors determined the underlying technical concern was a performance deficiency associated with the Mitigating Systems cornerstone that was more than minor because, if left uncorrected, would become a more significant safety concern. The underlying technical concern screened as a finding with very-low safety significance (Green) because, although it affected the design or qualification of the CVCS, it did not result in the loss of functionality of the CVCS. The violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The violation was categorized as a SL IV in accordance with Section 6.1.d.2 of the NRC Enforcement Policy because the changes were evaluated by the SDP, described above, as having very-low safety significance (i.e., Green finding). The inspectors did not identify a cross-cutting aspect associated with the finding because the finding was not representative of current performance.
05000255/FIN-2015004-032015Q4PalisadesFailure to Provide Bases to Determine Changes Did Not Involve Unreviewed Safety QuestionsThe inspectors identified a Severity Level (SL) IV, NCV of 10 CFR, Part 50, Section 59, Changes, Tests, and Experiments, for the licensees failure to maintain records of written safety evaluations, which provide the bases for concluding the nonsafety-related portions of the CCW system inside containment could be credited to perform their function during and following a DBE, and that the change would not result in an unreviewed safety question. The licensee entered this issue into their CAP and, after performing operability determinations, concluded the system would still be capable of performing its function. The violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The violation was categorized as a SL IV in accordance with Section 6.1.d.2 of the NRC Enforcement Policy because the resulting changes were evaluated by the SDP as having very-low safety significance (i.e., green finding). The resulting changes, the violations underlying technical concerns, impacted the Mitigating Systems cornerstone, and were evaluated separately as the Green finding with the associated 10 CFR, Part 50, Appendix B, Criterion II, NCV discussed above. The inspectors did not identify a cross-cutting aspect because cross-cutting aspects are not assigned to traditional enforcement violations.
05000255/FIN-2015004-022015Q4PalisadesFailure to Identify Components Required to be Covered by the Quality Assurance ProgramThe inspectors identified a finding of very-low safety significance, and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion II, Quality Assurance Program, for the licensees failure to identify all component cooling water (CCW) structures, systems, and components (SSC), which were required to be covered by the Quality Assurance Program (i.e., be safety-related). As a result, the licensee incorrectly credited nonsafety-related CCW components to remain functional during and following a design basis event (DBE). The licensee entered this finding into their CAP and, after performing operability determinations, concluded the system would still be capable of performing its function. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as having very-low safety significance (Green) because, although it was a deficiency affecting the design or qualification of a mitigating SSC, the SSC maintained its operability. The inspectors did not identify a cross-cutting aspect associated with this finding because it was determined not to be representative of current performance.
05000255/FIN-2015004-012015Q4PalisadesInadequate Dye Penetrant Examination of Pipe Lug WeldsThe inspectors 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 IX, Control of Special Processes, for the licensees failure to perform a dye penetrant (PT) examination of the Safety Injection System (SIS) pipe lug welds in accordance with the American Society of Mechanical Engineers (ASME) Code Section XI requirements. The licensee entered this issue into the Corrective Action Program (CAP) as CR-PLP-2015-04191, repeated the PT examination of the affected SIS lug welds to meet the full extent of coverage required by the ASME Code, repeated examinations of other welds conducted by the PT examiner during the outage, and removed the PT examiner from further weld examination activities. This performance deficiency was determined to be more than minor because, if left uncorrected, the failure to perform a PT examination in accordance with the ASME Code requirements could result in acceptance and return to service of a component with an undetected crack that would increase the possibility of pipe leakage or failure. In addition, the failure to perform a PT examination in accordance with the ASME Code adversely affected the Mitigating System Cornerstone attribute of Equipment Performance, because it could result in failure to detect cracks in pipe welds, which would reduce the availability and reliability of the SIS mitigating system. The inspectors evaluated the finding in accordance with IMC 0609, Appendix A, The SDP for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, and answered yes to screening question number 1. Although this finding adversely affected the design or qualification of the SIS pipe lugs, the finding screened as very-low safety significance (Green), because it did not result in the loss of operability or functionality of the affected SIS pipe segment. This finding had a cross-cutting aspect in the Field Presence component of the Human Performance cross-cutting area. Specifically, licensee leaders were not observed in the work areas of the plant to coach and reinforce standards or expectations for the licensees vendor staff to ensure deviation from standards and expectations were promptly corrected (H.2).
05000255/FIN-2015001-032015Q1PalisadesTurbine-Driven AFW Pump Trip During Surveillance TestingOn November 14, 2014, during performance of surveillance procedure RO-127, the turbine-driven AFW pump overspeed trip mechanism actuated and tripped the pump, which resulted in entry into 72-hour TS LCO 3.7.5 Condition A. Personnel performing the test reported no abnormalities prior to the trip. The pump had been operated satisfactorily on November 13, 2014, during performance of quarterly surveillance procedure QO-21, Inservice Test ProcedureAuxiliary Feedwater Pumps, Revision 43. Following the November 14 pump trip, the licensee performed a sequence of testing in an attempt to replicate the overspeed trip to assist in cause identification, or to demonstrate operability if the trip could not be replicated. This sequence included test runs in accordance with procedure SOP12, Feedwater System, Revision 72; procedure RO145, Comprehensive Pump Test Procedure, Auxiliary Feedwater Pumps P8A, P8B and P8C, Revision 13; and a modified test run in accordance with procedure T186, Auxiliary Feedwater Turbine K8 Overspeed Trip Test and Governor Setting, Revision 18. Since all acceptance criteria were met, the licensee declared the turbine-driven AFW pump operable on November 16, 2014, and LCO 3.7.5 was exited. The licensee subsequently conducted an ACE that identified three possible causes: 1) excess condensate in the moisture removal system, 2) decreased margin between the as-found operating speed and the overspeed trip setpoint, and 3) inherent design conditions of the steam supply and steam control system affecting the speed of the turbine. The evaluation also identified discrepancies in the Maintenance Rule classification of the steam traps and the frequency of preventive maintenance activities that were conducted on the equipment. Corrective actions were generated to evaluate each of the possible causes during the next pump maintenance outage, which occurred from March 26, 2015. The licensee conducted procedure T186 to verify the overspeed trip mechanism would trip within its required band and that the governor was functioning properly to changes in steam supply and pump speed. Procedure RO145 was conducted to observe steam trap operation and any oscillations in the steam supply system. Finally, the overspeed trip setpoint was revised based on the vendors recommendation for the operating conditions of this particular pump. At the end of the inspection period, all of the data from these tests were under review by the licensee, with vendor support, to aid in determining the cause of the November 2014 pump trip. Pending inspector review of the licensees data, conclusions, and any revisions made to the ACE, this issue is unresolved.
05000255/FIN-2014008-122014Q4PalisadesComponent Cooling Water System Licensing BasesThe inspectors identified an Unresolved Item (URI) regarding the licensing bases for the Component Cooling Water (CCW) system. Specifically, the inspectors require clarification as to what failures of the CCW system the licensee needs to postulate and evaluate. The NRC will conduct further inspection to determine when these changes to the licensing bases occurred. As part of the 2014 Component Design Bases Inspection (CDBI), the inspectors selected CCW pump P-52B and relief valve RV-0956 for review. Both of these components were part of the CCW system. The CCW system was designed as a closed cycle system, where both trains share a common suction and common discharge header. This means that although there were redundant pumps and heat exchangers, the system's piping was not designed to be redundant and a single pipe break or failure of the pressure boundary could result in the complete loss of CCW. One of CCW's safety functions was to transfer heat from the reactor and containment (post-Design Bases Events/Accidents) to the ultimate heat sink. Another important safety function for CCW was to provide cooling to the Engineered Safeguard Systems' (ESS) and containment spray (CS) pumps. Per the licensees design bases, cooling to the ESS pumps was required to maintain their operability. When reviewing the licensing bases for the plant, it was not clear what type of failures needed to postulated for the CCW system under post-accident conditions. The licensee's position was postulating a passive failure of CCW concurrent with a design bases accident (DBA) was not within their licensing bases. The licensee's position was that no active single failure, according to their definition in FSAR Section 1.4.16, would render CCW inoperable. They also considered a postulated failure of the non-safetyrelated portion of the CCW system inside containment as beyond design bases, except as result of a seismic event which was not postulated to occur in conjunction with an accident. Currently, the licensee credits post-accident heat being removed from containment by a combination of containment air coolers (CAC) and the containment spray (CS) system. The CAC are supplied by service water and are independent of the CCW system. Per the current design, the licensee needs either two CS pumps or one CS pump and three CACs. Both alternatives require the CCW system to remove heat from the CS system. However, the original design took credit for the CS and the CAC as independent and redundant in their capability to remove heat from the containment. In other words, originally the licensee needed either two CS pumps or three CACs. Additionally, the original design allowed for the capability to swap cooling water to the ESS pumps from CCW to service water remotely from the main control room (MCR). Both of these design flexibilities have been either lost or eliminated due to subsequent design changes. The inspectors noted the agency staff had previously evaluated the susceptibility of CCW to loss of function following certain assumed CCW pipe breaks during the Systematic Evaluation Program(SEP). This was documented on SEP Topic IX-3, Station Service and Cooling Water Systems Palisades, February 22, 1982. The agency staff had concluded the CCW design was not in conformance with GDC 44, regarding capability and redundancy of essential functions of the system. However, the staff noted the essential functions of CCW could be performed by other systems under all operating conditions. The SEP evaluation explicitly addressed a passive failure of the CCW system under post-accident conditions and concluded that the CACs would be capable of removing heat from containment. The inspectors were concerned that if the CCW system became inoperable as the result of non-safety-related component failures, the plant would no longer have the redundant capability to remove heat from the containment during a DBA, or provide alternate cooling to the ESS pumps from the MCR. In addition, the inspectors needed to clarify the licensing bases regarding a postulated loss of CCW concurrent with a design bases accident. This issue is unresolved pending further inspection to determine when these changes to the licensing bases occurred.
05000255/FIN-2014008-112014Q4PalisadesClassification of CCW Piping and Components Inside the Reactor Containment BuildingThe inspectors identified an unresolved issue (URI) regarding the Inservice Inspection (ISI) classification of component cooling water (CCW) piping and components inside the reactor containment building. This piping is currently classified as non-safety related. Resolution of this issue will be based on clarification of Palisades licensing basis by NRC staff. The inspectors reviewed SEP-ISI-PLP-002, ASME Code Boundaries for ASME Section XI Inservice Inspection Program, Revision 1. The purpose of this document was to establish classifications to ensure the proper scope of ASME Section XI system pressure tests, ASME Section V nondestructive examinations, ASME Section XI inspection of component supports, and ASME Section XI repairs, replacements and modifications. This program document stated that guidance for Class 2, 3, and nonsafety- related components is found in Regulatory Guide 1.26. The program referred to a set of color-coded P&IDs. The inspectors noted the CCW piping and components inside the reactor containment building were identified as non-safety-related on the color-coded P&IDs. The team also observed that Attachment 1 of SEP-ISI-PLP-002 (page 45 of 75) identified the piping from the containment penetrations to the single containment isolation valve outside containment as class 2 (check valve CK-CC910 for penetration MZ-14, air-operated valve CV-0911 for penetration MZ-15). This attachment referred to note H, which stated, in part: All containment penetration assemblies in Class 2, Class 3, and non-class piping will be considered ASME Class 2 out to the second isolation valve where applicable. In some cases, there is single isolation (e.g., MSS, FWS)... Where a Class 3 system penetrates containment, that portion will be considered Class 2 and treated as such (ie, ASME Section XI Interpretation BC84-603). The basis for classification of containment penetrations is contained in EGAD-EP-12, Mechanical Containment Penetrations Basis Program. The current Mechanical Containment Penetrations Basis document was SEP-APJ-PLP- 101, Revision 0. This document included a similar description for CCW penetrations MZ-14 and MZ-15. It stated that FSAR Table 5.8-4 classifies these penetrations as class C1. The FSAR table also identified these penetrations as Class C1 and indicated that they are not subject to 10 CFR Part 50, Appendix J testing requirements. The FSAR Section 6.7 included a description of Class C1 penetrations, it stated, in part: Penetrations in this class include those systems that are not connected to either the containment atmosphere or to the Primary Coolant System and are normally open or may be opened during power operation. These lines are protected from missiles originating inside the containment and the lines themselves form the boundary of the containment. One remote manually operated valve, locked closed manual valve or automatic isolation valve is provided in each line. Check valves are considered automatic. In SEP-APJ-PLP-101 also stated, that both Penetrations MZ-14 and MZ-15 were originally classified as C2 (closed system outside containment per FSAR Section 6.7), until FSAR Revision 22. These were reclassified by SDR-99-0884, dated July 22, 1999. The basis for these changes included an evaluation that determined CCW piping inside containment would not be damaged by internal missiles (as addressed by NRCs evaluation of SEP Topic III-4,c, dated September 21, 1981), and the classification of CCW as a closed system inside containment. The associated 10 CFR 50.59 evaluation determined that NRC approval was not required for these changes. The inspection team raised questions regarding the licensing basis classification of the CCW system inside containment: 1. FSAR Section 6.9 stated that Regulatory Guide 1.26 was used to select ASME Classes 2 and 3 systems and components for coverage by the inspection plan. Regulatory Guide 1.26, stated in part: The Quality Group C standards given in Table 1 of this guide should be applied to water-, steam-, and radioactive-waste-containing pressure vessels; heat exchangers (other than turbines and condensers); storage tanks; piping; pumps; and valves that are not part of the reactor coolant pressure boundary or included in Quality Group B but part of the following: ...(b) cooling water and seal water systems or portions of those systems important to safety that are designed for the functioning of components and systems important to safety, such as reactor coolant pumps, diesels, and the control room... The classification of the CCW system inside containment did not appear to be consistent with the guidance of Regulatory Guide 1.26. In addition, industry guidance provided by ANSI/ANS 51.1 1983 indicated that closed systems inside containment (with single active isolation valves) should be class 2 or class 3. 2. FSAR Section 6.7 addressed class C1 containment penetrations stating, These lines are protected from missiles originating inside the containment and the lines themselves form the boundary of the containment. The inspectors questioned whether it was appropriate to classify a portion of the containment boundary as nonsafety- related for the purpose of inspection, testing, and repairs. The licensee has stated that the CCW system classification is in accordance with their current licensing basis. This position is primarily based on the NRCs evaluation of SEP Topic III-4,c, dated September 21, 1981, the CCW system being protected from internal missiles, and seismic events not being postulated to occur coincidently with a LOCA. Resolution of this issue will be based on clarification of Palisades licensing basis. Pending resolution, this item will be tracked as an URI.
05000254/FIN-2013003-042013Q2Quad CitiesQuestion Concerning Availability of Dam Following a Seismic EventThe inspectors identified an unresolved item (URI) concerning the assumed availability of Lock and Dam No.14 following a design bases earthquake event. In a letter dated November 6, 1970 to the U.S. Atomic Energy Commission (now NRC), Commonwealth Edison (the licensee) addressed questions regarding the capability of the intake flume to withstand a seismic event. Specifically, the question stated: Demonstrate that the intake flume for plant cooling water either meets the seismic design requirements for a Class I structure or cannot fail in such a manner as to isolate the plant from the river cooling water source (emphasis added) in the event of a design basis earthquake. In response to Question 2.7, the licensee stated the retaining wall structure would remain intact during an operational bases earthquake and design bases earthquake. In addition, the licensee stated the earth embankment was found capable of resisting the sliding effects during a DBE. Lastly, the licensee stated the crib house would not fail and isolate the plant from the river water source. The design basis earthquake (DBE) at Quad Cities is 0.24g maximum horizontal ground acceleration coupled with other appropriate loadings to provide for containment and safe shutdown. The operating basis earthquake (OBE) is 0.12 g maximum horizontal ground acceleration. In 1998, the licensee contracted Ashton Engineering to determine the failure modes of Lock and Dam No. 14. In April 1998, Ashton Engineering provided the licensee with its conclusions, as documented in Study of Mississippi River Water Stage at Quad Cities Nuclear Power Station for Commonwealth Edison Company. The report cites the dam is located in Uniform Building Code (UBC) Seismic Zone 0 and therefore, design guidelines for the dam do not require evaluation for postulated earthquake loadings. The report does conclude the most likely damage during a seismic event would be a loss of the dam gate operating capability; however, the magnitude of the seismic event is not cited. Question 2.7 implies the river is considered available during a DBE event even though the downstream dam is not designed or constructed to remain functional during the assumed DBE. Although the site appears to be within their licensing bases (assume availability of the river during a DBE), the inspectors questioned whether this assumption considered actual potential consequences, i.e., the need to assume a loss of dam during a seismic event. This concern is considered an Unresolved Item (URI 5000254/2013003-04; 05000265/2013003-04, Question Concerning Availability of Dam Following a Seismic Event.) pending further consultation with the Office of Nuclear Reactor Regulation.
05000254/FIN-2013003-022013Q2Quad CitiesQuestion Concerning Licensing Bases of the Ultimate Heat SinkThe inspectors identified an unresolved item (URI) concerning the current licensing bases with respect to failure of Lock and Dam No. 14 on the Mississippi River. The inspectors reviewed several documents to ascertain the current licensing bases for the UHS. The river serves as a source of raw water for the station as well as one of the heat sinks. The licensee designated the UHS as non-safety related and as described in UFSAR Sections 2.4.4 and 9.2.5.3, the UHS is required if the river can no longer support its functions. The inspectors noted the current UFSAR states the loss of river results from damage to the Lock; however, historical documents state the event promulgates from a loss of Dam No. 14. Specifically: Section 2.4 of the Final Safety Analysis Report (FSAR) states, in part, the following: The river level at the station is assumed to drop to elevation 561 feet 0 inches if Dam No. 14 were to fail (emphasis added). Elevation 561 feet, 0 inches is the normal river level downstream of Dam 14. The station design includes the feature that at the time that Dam 14 fails (emphasis added) the only systems requiring the use of river water would be the RHR service water pumps. If in the unlikely event the broken dam condition occurs (emphasis added), it is necessary to open the gate on the ice melting line to permit the discharged water to return to the intake flume. This procedure permits the use of the water impounded in the intake flume and discharge flume to be used as an evaporative heat sink. This technique will impound 3,960,000 gallons of water. The maximum amount of water required by the system at this stage will be approximately 7000 gpm, which means that without any recirculation, the impounded water will last a minimum of 9.4 hours. It will be necessary to make up water to the intake flume under this condition by portable pumping equipment which will move water from the main river channel to the plant. Pumps capable of pumping 7000 gpm water will be maintained on site and backup pumps available from the local fire stations. In a letter dated November 6, 1970, to the U.S. Atomic Energy Commission (now NRC), Commonwealth Edison (the licensee) addressed questions regarding the use of portable equipment to move water from the main river channel to the plant under frozen river conditions. In response to Question 2.8, the licensee stated the portable pumps would not be required under these conditions. In addition, the licensee afforded the opportunity to clarify some statements in Section 2.4 of the Final Safety Analysis Report (FSAR). Specifically, the licensee stated the portable pumps would not be needed for makeup under a loss of dam event because the evaporative losses were about 20 gpm, not 7000 gpm as written. Although not stated in the licensees response, the inspectors noted the original description in FSAR Section 2.4 did not account for the ice melt line used to recirculate water back to the UHS; hence stating the need for makeup to the closed volume of the UHS. The licensees clarification (the need for 20 gpm makeup) accurately reflects the actual losses that need to be replenished. In the safety evaluation dated August 25, 1971, Section 2.3 states, the facility is also designed to provide an adequate supply of cooling water to the plant by providing a reservoir of about 3.8 million gallons of water in the intake bay so that, even if the river level dropped below a level of 565 feet MSL due to an assumed failure of Dam 14 downstream of the site (emphasis added), the water trapped in the intake bay would supply an adequate source of water for safe cool-down of the reactor primary system. In 1989, the licensee determined the original evaporative losses cited in the November 6, 1970, letter did not account for the worst case conditions. The licensee performed a calculation assuming worst case summer conditions and determined the evaporative losses were about 54 gpm. The licensee concluded the two 2000 gpm portable pumps were more than sufficient to address these losses. In NRC Inspection Report 05000254/1998-201, 05000265/1998-201(ML9805180380), the NRC team identified errors with the 1989 calculation and the licensees approach. Specifically, the team determined that in an evaporative mode, the trapped volume of UHS would increase in temperature and during summer operation, could be driven well above the 95 degrees Fahrenheit design temperature established for the RHRSW system. The licensee performed a preliminary calculation assuming 1 hour cool-down time on the main condensers from dam failure to loss of contact with the river (emphasis added) and determined the UHS could reach 112 degrees Fahrenheit. The team noted this evaluation used a method of makeup different than the current UFSAR. The team initiated an Unresolved Item 05000254/1998-201-12, 05000265/1998-201-12 to determine the resolution of the dam failure effects on the UHS. In May 1998, the licensee completed 10 CFR 50.59 Safety Evaluation SE-98-068. The purpose of this evaluation was to assess proposed changes to the UFSAR to incorporate the results of the study and revised temperature calculations. It described the change from dam failure to Lock and Dam failure as a clarification of the event to include a timeline and credible failure modes for the Lock and Dam. The licensee concluded these changes did not constitute an un-reviewed safety question, was not a change to a license condition and did not require a TS change. In May 1998, the licensee revised the UFSAR to reference the study and include details on the expected timeline and actions associated with a transportation accident impacting the Lock. The revised UFSAR also reiterates the portable pumps are onsite with backup pumps provided from another facility or leasing facility. In November 1998, an inspection was conducted to follow up on the licensees actions to address this URI. As documented in NRC Inspection Report 05000254/1998-019, 05000265/1998-019 (ML9812290045), the NRC team noted the licensee performed a hydraulic study of the Mississippi River in April 1998. This study assessed the possible failure modes of Dam No.14 and concluded the most credible and reasonable worst case scenario involved a transportation accident whereby a river barge impacts the Lock and Dam. The study determined the time for the river to separate from the UHS was about 90 hours. The licensee used this information in Calculation QDC-3900-M-0692 to determine the cooling needs for the UHS. The calculation concluded three portable pumps delivering a total of 5100 gpm of cooler water were needed to ensure the inlet temperatures remained within design limits. A violation for the failure to assure the design basis information was consistent with actual plant design was issued. As described in Section 1R07.1b(3), in April 2001, the licensee completed a 10 CFR 50.59 Safety Evaluation Screening, QC-S-2001-0026, to assess removal of the portable pumps from onsite and relocating the pumps to an offsite leasing facility located a few hours away. The licensee revised the UFSAR and removed the portable pumps from the site. The inspectors noted the original FSAR did not provide the detail as to the cause of the dam failure or a time line for the loss of river event. The licensee stated their response to Question 2.8, (and as reiterated in the August 1971 Safety Evaluation) implies the loss is not immediate because water level would recede in the condenser box and the unit would be shutdown due to loss of condenser vacuum. The licensee contends the main condenser would remain functional following a dam failure up until the vacuum can no longer be maintained by the UHS supply. The inspectors noted the licensee had assumed 1-hour of such operation in their preliminary calculation performed in November 1998. The inspectors were concerned the licensee redefined the loss of river event from the original Section 2.4 of the FSAR description of an unlikely event of a broken dam to a transportation accident impacting the lock. Therefore, this issue is considered an Unresolved Item (URI 5000254/2013003-02; 05000265/2013003-02, Question Concerning Licensing Bases of the Ultimate Heat Sink) pending further consultation with the Office of Nuclear Reactor Regulation.
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