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05000440/FIN-2018003-012018Q3PerryApplication of ASME Code Case N5133 for the Emergency Service Water Piping DegradationsThe inspectors identified an Unresolved Item concerning the Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B, and ASME Code requirements for the ESW piping systems with regards to the licensees application of ASME Code Case N5133, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping, Section XI, Division 1. Updated Safety Analysis Report (USAR) Section 9.2.1 describes that the function of ESW system is to provide a reliable source of water to safety-related components required for normal and emergency reactor operation. USAR Table 3.21, Equipment Classification, delineates that the ESW piping system is safety-related and designed in accordance with the requirements of ASME Section III, Subsection ND (Class 3). The regulation in 10 CFR 50.55a(g) requires, in part, that Class 3 components and their supports meet the requirements of ASME Section XI of the ASME Boiler and Pressure Vessel (BPV) Code or equivalent quality standards. The ASME also publishes Code Cases, which provide alternatives to existing Code requirements. The NRC Regulatory Guide (RG) 1.147 identifies that Code Case N5133 provides acceptable alternatives to applicable parts of Section XI, provided it is used with any identified conditions or limitations. Code Case N5133, Section 2(d) requires that a flaw evaluation shall be performed to determine the conditions for flaw acceptance. Section 3 provides accepted methods for conducting the required analysis. In addition, Section 3 requires, in part, that nonplanar flaws shall be evaluated in accordance with the requirements in 3.2. Additionally, Section 5 requires that an augmented volumetric examination or physical measurement to assess degradation of the affected system shall be performed as follows: (a) From an engineering evaluation, the most susceptible locations shall be identified. A sample size of at least five of the most susceptible and accessible locations, or, if fewer than five, all susceptible and accessible locations shall be examined within 30 days of detecting the flaw. (b) When a flaw is detected, an additional sample of the same size as defined in 5(a) shall be examined. (c) This process shall be repeated within 15 days for each successive sample, until no significant flaw is detected or until 100 percent of susceptible and accessible locations have been examined. On June 13, 2018, a through-wall leakage on the 20 ESW piping was identified in CR 201805504. As a result, the licensee invoked the Code Case to evaluate this flaw and permit the degraded ESW piping system to remain in service for a limited period without repair/replacement. The licensees evaluation involved characterization of this flaw as nonplanar, and subsequently, the methodology as described in Section 3.2 of the Code Case was utilized for this nonplanar flaw. Additionally, the licensee identified the five most susceptible and accessible locations in the ESW system and performed examination in accordance with Section 5(a). From the examination of the five additional locations, another localized wall degradation was identified on the 8 ESW pipe elbow on July 10, 2018. The licensee initiated CR 201806205 to document this condition. The licensee characterized this degradation also as a nonplanar flaw, and this degradation represented approximately 80 percent wall loss from its nominal thickness. During the review of the licensee evaluation of this degraded pipe elbow, the inspectors identified that the methodology as described in Section 3.2 of the Code Case had not been utilized. Instead, the licensee elected to use an alternate methodology to evaluate and disposition for its acceptability. Furthermore, the inspectors identified that the licensee essentially redefined the term flaw in the Code Case to reflect the ASME Section XI, IWA9000 definition of the term defect. The ASME Section XI, IWA9000 defines a flaw as an imperfection or unintentional discontinuity that is detectable by nondestructive examination. It also defines a defect as a flaw (imperfection or unintentional discontinuity) of such size, shape, orientation, location, or properties as to be rejectable. With respect to the Code Case, the licensee essentially restricted the criteria for examination scope expansion only to the flaws that were rejectable; therefore, the licensee had not expanded the scope to perform examination of additional locations in accordance with Section 5(b). In essence, two items are to be further evaluated and addressed: (1) whether the use of methodology not described in the Code Case Section 3.2 was appropriate for evaluation of the nonplanar flaw on the 8 ESW pipe elbow, and (2) whether the stopping of scope expansion for examination as required by the Code Case Section 5(b) was appropriate based on the licensees redefining of the term flaw. In response to the inspectors concern, the licensee initiated CR 201808483, NRC ID: Code Case N5133 Interpretation, September 26, 2018. The licensee also plans to perform examination of five additional locations in November of 2018. This represents an item where the inspectors identified Code interpretation issues that resulted in a disagreement with the licensee. This will require additional review to determine whether a violation exists. Therefore, this issue is considered an unresolved item pending completion of inspector review and evaluation and discussion with the Office of Nuclear Reactor Regulation. Licensee Action: The licensee plans to perform examination of five additional locations in November of 2018. Corrective Action Reference: CR 201808483
05000331/FIN-2018003-012018Q3Duane ArnoldLicensee-Identified ViolationA violation of very low safety significance (Green)was identified by the licensee and has been entered into the corrective action program. This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Title 10 of the Code of Federal Regulations (CFR) 50, Appendix B, Criterion III, states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. System Design Specification APEDA61019, Pressure Integrity of Piping and Equipment Pressure Parts Data Sheet, required in the applicable castings section T1.3.3.b, all accessible surfaces including machine surfaces shall be examined by either the magnetic particle or liquid penetrant method in either the furnished or finished condition. Contrary to the above, in October 2016, measures were not established to assure that applicable design basis requirements as defined in 10 CFR 50.2 were translated into work instructions repairing the B inboard main steam isolation valve, CV 4415, during RFO 25. Specifically, instructions to perform a NDE of machined surfaces following the valve repair were not included in the work package. As a result, the non-destructive examination was not performed prior to placing the valve into service.
05000331/FIN-2018003-022018Q3Duane ArnoldMinor ViolationDuring Mode 1 power operations on July 9, 2018, the licensee had both doors of a secondary containment airlock open simultaneously, and a minor violation of Technical Specification (TS) 3.6.4.1 Secondary Containment was self-revealed. During the time both doors were open, approximately 3 seconds, the allowable penetration opening area was exceeded and rendered the secondary containment inoperable. Technical Specification 3.6.4.1 requires secondary containment to be operable in Modes 1, 2 and 3. Technical Specification Surveillance Requirement 3.6.4.1.2 supports secondary containment operability by verifying that either the outer door(s) or the inner door(s) in each secondary containment access opening are closed. The posted instructions at each secondary containment airlock door stated, ATTENTION Push Button To Be Held In For 2 Seconds Prior To Opening Door, to be of a type appropriate for traversing the containment airlock. Contrary to the above, at approximately 1:34 p.m. on July 9, 2018, while operating in Mode 1 at 97 percent power, two individuals simultaneously traversing through opposite doors of a secondary containment airlock each failed to hold the airlock interlock push button for two seconds prior to opening their respective doors resulting in a momentarily inoperability of secondary containment. Operability was restored upon the immediate closure of one of the two doors. Subsequently, maintenance was unable to recreate the condition and satisfactorily performed Surveillance Test Procedure (STP) 3.6.4.102, Secondary Containment Airlock Verification, and GMPELEC44,Section A5.1,Airlock Door Interlock Checks.The licensee entered this
05000315/FIN-2018002-062018Q2CookMinor ViolationTechnical Specification (TS) 5.4, Procedures, requires that the applicable procedures recommended in Regulatory Guide 1.33 be established, implemented, and maintained. Regulatory Guide 1.33 states that maintenance that could affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with procedures appropriate to the circumstances. Contrary to this requirement, procedure 12EHP4030056218, Automatic Operation of Auxiliary Feedwater Pumps, was not performed as written in the procedure. Specifically, pages were skipped which resulted in the 2CD EDG inadvertently starting during the surveillance. Screening: The issue resulted in momentary loss of the T21C and T21D vital busses until the 2CD EDG reached rated speed and connected to the busses. The reactor was defueled at the time. One train of spent fuel pool cooling was lost for several minutes, but the other train stayed in service and there was no apparent change in spent fuel pool temperature. The issue screened as minor based on the guidance in IMC 0612 Appendix E because there were no safety consequences and there was no transient of any significance. Violation: This failure to comply with TS 5.4 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000315/FIN-2018002-052018Q2CookMinor ViolationWhile there did appear to be a reduction in operational errors being made in the field while manipulating equipment (such as during clearance activities and in performing certain evolutions) the inspectors noted a trend in configuration control issues. Most of these dealt with some kind of operation department interface or coordination with another department. In one case, valves associated with feedwater heater level control were left closed following a project to replace some of the heaters, which contributed to a manual reactor trip due to high moisture-separator drain tank level when starting the plant following the Unit 2 refueling outage. Other examples were Chemistry and Operations department coordination on an non-essential service water (NESW) valve alignment which led to NESW being isolated to generator seal oil cooling during plant startup, poor coordination between Maintenance and Operations which resulted in a containment penetration being left open, a pressure gauge remaining isolated after the Projects department completed the heater drain pump replacements, and the failure to ensure that valve-closure tests were done following the feedwater heater replacements. Another identified trend was in the area of post-maintenance testing (PMT). During the refueling outage on Unit 2, both the NRC and the licensee identified instances of improper PMTs being scheduled for safety-related equipment. Inspectors identified work on an EDG fuel oil transfer pump that did not have an in-service test (IST) scheduled. The licensee identified the lack of a time response test following a motor-driven AFW pump motor replacement, was a repeat issue from the previous outage. The licensee also identified the lack of an IST following a seal replacement on a CCW pump. In each case, the issues were discovered and corrected before equipment was restored to fully operable status. In response to the trend, the licensee reviewed other work on safety-related equipment for the outage to confirm the proper PMTs would be done. No other issues were identified. Finally, early in the observation period, the inspectors noted a trend in procedure quality for maintenance activities on safety-related equipment. There were instances regarding Turbine-Driven Auxiliary Feedwater (TDAFW) pump linkages where better procedure direction could have precluded binding and governor-valve travel issues. Additionally, while replacing a TDAFW governor, a snap ring was inadvertently left out of a coupling due to insufficient procedure detail. Regarding the EDGs, the licensee discovered instructions for assembly of air start check valves did not contain the torque guidance that the vendor drawings stipulated. In response to this trend, the licensee started to perform deliberate reviews of OE before maintenance on some safety-related equipment, to verify maintenance instructions had up-to-date guidance before starting work. No violations or findings were identified by the inspectors. 12 Licensee management acknowledged the issues discussed by the inspectors.
05000315/FIN-2018002-042018Q2CookLicensee-Identified ViolationThis violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Enforcement: Violation: Title 10 Code of Federal Regulations; Part 20.1501(c) requires that the licensee shall ensure that instruments and equipment used for quantitative radiation measurements are calibrated periodically for the radiation measured. Contrary to the above, between November 2012 and May 2017 the licensee used the liquid scintillation counter for quantitative radiation measurements outside the range of equipment capability and the system calibration. The licensee analyzed the impact on the annual effluent reports and UFSAR limits between 1/8/2013 and 5/3/2017. The licensee entered the violation on the corrective action program. Licensee Identified Non-Cited Violation Significance/Severity Level: Green. The inspectors determined the performance deficiency was more than minor because it adversely affected the Plant Facilities/Equipment and Instrumentation attribute of the Public Radiation Safety Cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. The inspectors assessed the significance of the finding usingSDP Appendix D and concluded the violation was of very low safety or security significance (Green). Corrective Action Reference: AR20174835
05000373/FIN-2018002-042018Q2LaSalleMinor Violation - Follow-up of Events and Notices of Enforcement Discretion

Minor Violation: For S/RV 2B21F013L, serial number N63790050012 (hereafter referred to as S/RV 12), the licensee completed a work group evaluation as documented in AR 03975216ACIT No. 3 to investigate the cause for two S/RVs that failed a set pressure lift test out of specification low. For ACIT No. 3, the licensee staff incorporated a vendor letter that documented the results of the S/RV vendors review of the S/RV 12 condition and which recorded an out of tolerance spring condition. It stated that The spring was measured and rate tested. The free height was found to be below the minimum original equipment manufacturer specified tolerance. The licensees vendor subsequently replaced the nonconforming spring with a new spring. In prior vendor correspondence with the licensee (reference E-mail dated June 24, 2015), the vendor stated that Typically we contribute a low as-found lift to an out-of-tolerance spring rate or free height dimension. Therefore, the nonconforming spring free height dimension may have caused the low as-found lift setpoint failure for this valve and as such was relevant (e.g. material) to the determination of a failure cause that was reported in LER 05000374/201700400 and 01. However, the licensee failed to identify this during their cause investigation and erroneously reported in LER 05000374/201700400 and 01 that The vendor reported for both valves that all the spring tolerances were within the acceptance limits. The licensee documented this violation in AR 04134591, Potential Minor Violation for Unit 2 LER 20170401. The licensee also submitted a revision to the LER as LER 05000374/201700402 

Screening: The significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which could impede the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. The inspectors determined that this issue was a Severity Level IV violation based on Example 6.9.d.10 in the NRC Enforcement Policy which states, A failure to identify all applicable reporting codes on a Licensee Event Report that may impact the completeness or accuracy of other information (e.g. performance indicator data) submitted to the NRC. In accordance with the Section 2.2.1.c of the NRC enforcement policy, the severity level of a violation involving the failure to make a required report to the NRC will depend on the significance of and the circumstances surrounding the matter that should have been reported. The NRC had not relied on information in this LER report to make a regulatory decision, and the inspector answered no to each of the more than minor screening questions in Appendix B of IMC 0612 for the issue of concern. Therefore, the NRC determined this was a minor violation because it was associated with a minor performance deficiency. Violation: Failure to comply with 10 CFR 50.9 Completeness and accuracy of information and accurately report the nonconforming S/RV 12 spring tolerance in LER 05000374/201700400 and 01 to the NRC constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000315/FIN-2018002-032018Q2CookLicensee-Identified Violation

This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy 

Violation: Title 10 of the Code of Federal Regulations (10 CFR) 50.47 b(8) requires that licensee emergency plans meet the standard of having adequate emergency facilities. The Cook Plant Emergency Plan states that the Technical Support Center (TSC) (an emergency facility) will be constructed to provide the same degree of radiological habitability as the Control Room under accident conditions. Contrary to the above, from January 24 to 30, 2018, the licensee failed to maintain the TSC as an adequate emergency facility, by installing a portable air conditioning unit in the Shift Managers office which compromised the ability of the TSC ventilation system to fulfill its function of providing the necessary radiological protection for the TSC. Specifically, the exhaust from the portable unit was routed to an existing ventilation duct of the TSC ventilation system, and a panel on one of the ventilation units was opened, exposing the TSC to the turbine building environment. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it adversely affected the Facilities and Equipment attribute of the Emergency Preparedness cornerstone, whose objective is to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors assessed the significance of the finding usingSDP Appendix B and concluded the violation was of very low safety or security significance (Green). Corrective Action Reference: AR20180952
05000373/FIN-2018002-032018Q2LaSalleLicensee-Identified Violation

This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Technical Specification LCO 3.4.4 (applicable for Modes 1, 2 and 3) states: The safety function of 12 safety relief valves (S/RVs) shall be OPERABLE, and Action Statement A states that One or more required S/RVs inoperableA.1 be in mode 3 in 12 hours and A.2 be in Mode 4 in 36 hours. Technical Specification SR 3.4.4.1 states that Verify the safety function lift setpoints of the required S/RVs are as follows 

Number of S/RVs Setpoint (psig

2 1205 36.

3 1195 35.

2 1185 35.

4 1175 35.

2 1150 34.

Contrary to the above, during portions of previous Unit 1 and 2 operating cycles from 2012 through January of 2017, two main steam S/RVs did not meet these lift pressure setpoint requirements. Specifically S/RV 2B21F013C lifted at 1131 psig instead of from 1139.8 to 1210.2 psig and S/RV 2B21F013L lifted at 1130 psig instead of from 1159.2 to 1230.8 psig (reference: Licensee Event Report 05000374/201700400; 01, Two Main Safety Relief Valves Failed Inservice Lift Inspection Pressure Test.

Significance/Severity: This licensee identified finding affected the Initiating Events Cornerstone and was screened in accordance with Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At Power. The two affected SRVs lifted low outside of their setpoint band, which was conservative with respect to maintaining the reactor coolant system overpressure protection safety function of these valves. Therefore, the inspectors determined that this finding is of very low safety significance (Green) because after a reasonable assessment of degradation, the finding would not have resulted in exceeding the reactor coolant system leak rate for a small LOCA and did not affect other systems used to mitigate a loss-of-coolant accident. Corrective Action Reference: AR 3974669
05000315/FIN-2018002-022018Q2CookLicensee-Identified ViolationThis violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Enforcement: Violation: License conditions 2.C.(4) (Unit 1) and 2.C.(3)(o) (Unit 2) require implementation of the approved fire protection program. Per the Cook NFPA 805 Fire Protection Program Manual Sections 3.11.2 and 3.11.4, fire seals shall have at least a three hour fire rating. Contrary to the above, on February 6, 2018, the licensee identified multiple fire seals (many of which were between the control rooms and the cable spreading area underneath) that were degraded to the point that they could no longer meet the three hour rating requirement of Sections 3.11.2 and 3.11.4 of the Cook NFPA 805 Fire Protection Program Manual. Specifically, inadequate controls in the fire seal maintenance procedure and unclear guidance for Performance Verification department inspections led to a deterioration in seal quality. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it adversely affected the Protection Against External Factors attribute of the Mitigating Systems cornerstone, whose objective is to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). The inspectors assessed the significance of the finding usingSignificance Determination Process Appendix F and concluded the violation was of very low safety significance (Green).Corrective Action Reference: AR20181208
05000373/FIN-2018002-022018Q2LaSalleFailure to Follow Procedure and Perform Database Revision Review RequirementsThe inspectors identified a Green finding of very low safety significance for the licensees failure to follow procedure NSWPWM03, Predefine Database Revisions, Revision 0, for retiring procedure LESGM108, Inspection of 480V Motor Control Center Equipment, that performed bus bar inspection on Division 3 motor control centers. Specifically, instead of completing NSWPMW03, step 6.5, Database Revision Review Requirements, to retire the bus bar inspections for Division 3 motor control centers, the licensee retired the procedure based solely on having previously retiring the bus bar inspections for Division 1 and Division 2 in 2002,and did not performthe required review.
05000373/FIN-2018002-012018Q2LaSalleFailure to Implement a Preventative Maintenance Strategy for Residual Heat Removal Service Water Pump Shorting RelaysA self-revealed Green finding of very low safety significance was identified for the licensees failure to implement a preventative maintenance (PM) strategy for the residual heat removal service water (RHRSW) pump shorting relays in accordance with procedure MAAA716210, Performance Centered Maintenance (PCM) Process, Revision 11. Specifically, a PCM template was issued in 2002 that required periodic as-found testing and calibration for control and timing relays, but a maintenance strategy was never implemented. As a result, one of the normally closed contacts on the Unit 1 D RHRSW pump shorting relay developed a high contact resistance and prevented the Unit 1 D RHRSW pump from starting.
05000316/FIN-2018002-012018Q2CookSteam Dump Closure Caused by Human ErrorOn May 10, 2018, a Green self-revealed finding and associated Non-Cited Violation occurred when licensee personnel caused the Unit 2 steam dump valves to the condenser to close. Specifically, when tuning the controller for the steam dump valves, licensee personnel left the controller in automatic, resulting in the closure of all the steam dump valves. This caused both the steam generator power operated relief valves and a steam generator safety valve to lift.
05000266/FIN-2018001-052018Q1Point BeachLicensee-Identified ViolationViolation: Technical Specification (TS) 3.0.4 states in part that entry into a MODE or other specified condition in the Applicability of a limiting condition for operation (LCO) shall only be made when the LCOs Surveillances have been met... TS 3.7.5 Auxiliary Feedwater (AFW) Limiting Condition SR 3.7.5.1 required in part Verify each AFW manual, power operated, and automatic valve in each water path, and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position. Contrary to the above, at 1500 on October 29, 2017, Unit 1 entered MODE 3 and the licensee failed to verify that AFW (System required for MODE 3) turbine driven (TD) AFW steam supply valves 1MS235 and 1MS237 were in the correct (open) position. These valves were in fact shut rendering the TDAFW pump inoperable until the licensee identified this error and opened these valves at 1610 on October 29, 2017(reference; Licensee Event Report 05000266/201700200, Operation or Condition Prohibited by Technical Specifications). Significance/Severity: This licensee identified finding, affected the Mitigating Systems Cornerstone and was screened in accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At Power, issued June 19, 2012. Because of the short duration (~1 hour) that the TDAFW pump was not operable, the inspectors determined that this finding is of very low safety significance (Green) because: the performance deficiency was not a design or qualification issue; it did not represent a loss of the system function; the train was neither inoperable for greater than its allowed outage time nor was it inoperable for greater than 24 hours; and was not part of an external event mitigating system. Corrective Action Reference: AR 02233500 Made Mode Change With Inoperable TDAFW
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.
05000266/FIN-2018001-042018Q1Point BeachEnforcement Action: EA18030: Unanalyzed Condition for Tornado Generated MissilesOn 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 an 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 12 Point Beach, the EGM provided for enforcement discretion of up to three 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 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. Table 1.31 of the Point Beach Final Safety Analysis Report (FSAR) states in part that SSCs which are essential to the prevention and mitigation of nuclear accidents shall be designed, fabricated, and erected to withstand the forces that might reasonably be imposed by the occurrence of an extraordinary natural phenomenon such as a tornado. On March 1, 2018, the licensee initiated AR 02252240, identifying a nonconforming condition of Table 1.31. Specifically, on both units 1 and 2, the steam supply lines and exhaust stacks for the turbine-driven auxiliary feedwater pumps, the main steam isolation valves, the atmospheric steam dumps, the main steam safety valves, and the vents for T175B bulk fuel oil storage tank were not adequately protected from tornado-generated missiles. The licensee declared the affected SSCs inoperable and promptly implemented compensatory measures designed to reduce the likelihood of tornado-generated missile effects. The condition was reported to the NRC as Event Notice (EN) 53239 as an unanalyzed condition and potential loss of safety function. Corrective Actions: The licensee documented the inoperability of the SSCs and the affected TS LCO conditions in the CAP and in the control room operating log. The shift manager notified the NRC resident inspector of implementation of EGM 15002, and documented the implementation of the compensatory measures to establish the SSCs operable but nonconforming prior to expiration of the LCO required action. The licensees immediate compensatory measures included: review and revision of procedures for a tornado watch and a tornado warning to provide additional instructions for operators preparing for tornados and/or high winds, and a potential loss of SSCs vulnerable to the tornado missiles; confirmation of readiness of equipment and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX); verification that training was up to date for individuals responsible for implementing preparation and response procedures; and establishment of a heightened station awareness and preparedness relative to identified tornado missile vulnerabilities. Corrective Action Reference: AR 2252240 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); TS 3.7.1, Main Steam Safety Valves (MSSVs); TS 3.7.2, Main Steam Isolation Valves (MSIVs) and Non-Return Check Valves; TS 3.7.4, Atmospheric Dump Valve (ADV) Flowpaths; TS 3.7.5, Auxiliary Feedwater (AFW); TS 3.8.1; AC Sources - Operating; and TS 3.8.3, Diesel Fuel Oil and Starting Air. 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 15-002 because the licensee initiated initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. The licensee implemented actions to track the more comprehensive actions to resolve the nonconforming conditions within the required 60 days. These comprehensive actions are to remain in place until permanent repairs are completed, which for Point Beach were required to be completed by June 10, 2018, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC such that discretion was no longer needed
05000315/FIN-2018001-022018Q1CookOperation of Letdown System Leads to Voiding and Subsequent Relief Valve LiftThe inspectors identified a finding of very low safety significanceand associated Non-Cited Violation of Technical Specification 5.4, Procedures, when the licensee failed to maintain a procedure for operating the letdown system. As a result, a water-hammer occurred which caused a safety-related relief valve to lift, which discharged reactor coolant to the Pressurizer Relief Tank until letdown was isolated
05000315/FIN-2018001-012018Q1CookFailure of Unit 1 Turbine Driven Auxiliary Feedwater Pump to Reach Rated SpeedA self-revealed finding of very low safety significance with an associated Non-Cited Violation of Technical Specification 5.4 Procedures, occurred on December 21, 2017, when the Unit 1 Turbine-Driven Auxiliary Feedwater Pump failed to reach rated speed during a surveillance. Procedure 12MHP5021056008, Turbine-Driven Auxiliary Feedwater Pump Governor Valve Maintenance, was not appropriate for the circumstances in that direction was not given to check that the governor valve could fully open following maintenance on the governor valve.
05000266/FIN-2018001-032018Q1Point BeachInadequate Basis for Deletion of TRM 3.4.3 Primary System Integrity RequirementsThe inspectors identified a Severity Level IV, Non-Cited Violation of 10 CFR 50.59, Changes, Tests, and Experiments, and an associated finding of very low safety significance (Green) for failure to provide a written evaluation, which provided the basis for the determination that a change did not require a license amendment. Specifically, the licensee failed to provide a basis for why deletion of the nondestructive examination requirements in Technical Requirements Manual (TRM) 3.4.3 for Primary System Integrity did not require prior NRC approval.
05000266/FIN-2018001-022018Q1Point BeachFailure to Evaluate Material Acceptability for a Safety-Related DoorstopA self-revealed Green finding and associated Non-Cited Violation of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III was identified when the licensee failed to evaluate the suitability of material prior to installation in the plant. Specifically, the licensee installed a doorstop, which was fabricated from a length of Unistrut, behind a safety-related door. The Unistrut was not suitable for the application and caused the door to become wedged open.
05000266/FIN-2018001-012018Q1Point BeachFailure to Evaluate and Characterize Fire Protection Pipe Wall DegradationThe inspectors identified a finding of very low significance, for the failure to follow procedure NP 7.7.22, Service Water and Fire Protection Inspection Program. Specifically, Section 4.10, Degraded Component Characterization and System Failure Analysis, step 4.10.1 states, in part, the extent of pipe wall degradation shall be characterized by volumetric non-destructive examination (NDE) for subsequent flaw evaluation. The licensee identified pipe corrosion on November 28, 2012, and failed to characterize it by volumetric NDE.
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.
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-2018001-012018Q1Quad 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.
05000461/FIN-2017004-012017Q4ClintonReactor Down Power due to Reactor Recirculation Pump Motor Lower Bearing Oil LeakA self-revealed finding of very low safety significance was identified for the licensees failure to comply with the requirements of station procedure MAAA716004, Compression Fittings Inspection, Installation, Remake and Repair, Revision 3. Specifically, the licensee failed to properly assemble a joint that was part of the reactor recirculation (RR) pump motor B oil level monitoring system that subsequently leaked requiring the plant operators to perform an unplanned power reduction to allow for identification and repairs of the leak. The licensee documented this issue in the corrective action program (CAP) as Action Request (AR) 04029024. As corrective actions, the licensee made repairs to the effected joint, inspected the remaining joints to ensure proper integrity, and filled the lower bearing reservoir. This issue was more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and impacted 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 leak on the reactor recirculation pump motor oil level monitoring system would have eventually resulted in the failure of the reactor recirculation pump causing a transient and upsetting plant stability. This finding was determined to be of very low safety significance because the event did not cause a reactor trip. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of work management, where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, the licensee failed to ensure that all personnel installing the compression fittings received an installers brief prior to performing work on the reactor recirculation pump motor oil level monitoring system. (H.5)
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-2017004-022017Q4CookUnit 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-2017004-032017Q4CookFailure 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-2017004-042017Q4CookFailure 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.
05000461/FIN-2017004-022017Q4ClintonFailure to Assess and Manage Risk Associated with the Performance of Control Rod Time TestingThe inspectors identified a finding of very low safety significance and a non-cited violation of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to assess and manage the increase in risk that may result from proposed maintenance activities. Specifically, the licensee failed to assess and manage the increase in risk associated with performing control rod scram time testing in Mode 1, prior to performing the activity. As corrective actions, the licensee assessed the increase in risk for performing control rod scram time testing at power and developed a risk mitigation plan that was used to complete the testing. This performance deficiency was determined to be more than minor because the finding was associated with the human performance attribute of the Initiating Events cornerstone and impacted 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. Specifically, the failure to assess risk and develop a risk mitigation plan for control rod time testing at power contributed to an automatic reactor scram. Using IMC 0609, Attachment 4, Initial Characterization of Findings, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, issued May 5, 2005, and Appendix M, Significance Determination Process Using Qualitative Criteria dated April 12, 2002, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance. Specifically, the inspectors and the Region III Senior Reactor Analyst (SRA) determined that Appendix K was not directly applicable to this finding because the licensee performs qualitative evaluations of maintenance activities that have the potential to cause a transient rather than quantitative evaluations. The SRA used insights from Appendix K to support a qualitative SDP evaluation using the principles of Appendix M. The SRA determined that the maintenance activity could only result in an uncomplicated reactor transient event and that the increased risk of a transient compared to the baseline risk of the plant was of very low safety significance. The SRA considered the conditional core damage probability of an uncomplicated transient in this evaluation, which was less than 1E6, to conclude that the finding was Green. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of work management, where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. While performing the work order risk screening for completing the control rod scram time testing while the reactor was shut down, the screener identified that a new screening would be needed if the testing was performed at power. However, no holds were placed on the work order to ensure the risk screening was completed. (H.5)
05000461/FIN-2017004-032017Q4ClintonFailure to Identify the Extent of Condition for an Inadequate 10 CFR 50.59 EvaluationThe inspectors identified a finding of very low safety significance and a non-cited violation of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the licensees failure to follow a procedure that implemented the Quality Assurance Program requirements. Specifically, the licensee failed to follow procedure PIAA1251003, Corrective Action Program Evaluation Manual, and identify the extent of condition for a lack of proficiency in applying the licensing basis when performing 10 CFR 50.59 evaluations. The licensee documented this issue in their CAP as AR 04075581. The licensee planned an Updated Safety Analysis Report (USAR) Upgrade Project which reportedly would include a review of safety evaluations for USAR changes that dated back to 1986 and determined the scope of this project would be adequate to identify the extent of condition for this issue. The inspectors determined that this issue was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, because the extent of condition review was not adequate, there is a potential for other safety systems to have been adversely affected by a lack of proficiency in applying the licensing basis during safety related system changes. As a result, safety-related systems may not be able to perform intended safety functions as defined in the USAR. This issue would also adversely affect the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding was screened against all cornerstones and determined to be of very low safety significance because the finding met each of the applicable screening questions to be characterized as having very low safety significance. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of procedure adherence, which stated individuals follow processes, procedures, and work instructions. Specifically, when the NRC violation was documented in the CAP previously it was not appropriately classified in accordance with PIAA120 and was also incorrectly closed to an unrelated evaluation. This contributed to the failure to appropriately perform an extent of condition. (H.8)
05000461/FIN-2017004-042017Q4ClintonFailure to Perform an Evaluation in Accordance with 10 CFR 72.48 for Changes Made to the Time-to-Boil CalculationThe inspectors identified a Severity Level IV non-cited violation of 10 CFR 72.48(d)(1), Changes, Tests, and Experiments, for the licensees failure to perform a written evaluation which provides the bases for the determination that changes do not require a Certificate of Compliance amendment pursuant to 10 CFR 72.48(c)(2). Specifically, the licensee accepted Engineering Change Order ECO501825R0 (1), ECO501848R1 (1), and ECO501848R1 (4) on June 20, 2016, to the time-to-boil calculation as described in the HI-STORM FW Final Safety Analysis Report and incorrectly screened out performing an evaluation of those changes in accordance with 10 CFR 72.48. The licensee documented this issue in its CAP as AR 02714091 and AR 04081583. The licensee is performing a 10 CFR 72.48 evaluation for Engineering Change Order ECO501825R0 (1) and ECO501848R1 (4) while planning to revise the acceptance of ECO501848R1(1). The inspectors determined that the violation was of more than minor significance as the inspectors could not reasonably conclude that the above changes did not require prior NRC approval. The violation screened as a Severity Level IV non-cited violation using example 6.1.d.2 of the NRC Enforcement Policy. No cross-cutting aspect was identified since cross-cutting aspects are not assigned to traditional enforcement violations.
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.
05000456/FIN-2017002-032017Q2BraidwoodFailure to Adequately Implement and Maintain the Radiological Environmental Monitoring Program by Collecting Representative Samples from the Principal Food PathwaysGreen. A finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix I, Section IV(B), were identified by the inspectors for the licensees failure to establish an appropriate surveillance and monitoring program in order to provide data on measurable levels of radiation and radioactive materials in the environment to evaluate the relationship between quantities of radioactive material released in effluents and resultant radiation doses to individuals from principal pathways of exposure. This was an NRC-identified finding for the failure to implement and maintain the licensees radiological environmental monitoring program (REMP) by collecting representative samples from the highest deposition coefficient (D/Q) quadrant locations during annual REMP sampling and collections of food products in 2015. On May 25, 2016, during a review of the stations annual radiological environmental operating report for 2015, the inspectors noted that the licensee documented missed samples in three out of four quadrants where the principal food pathways were grown within the 10 kilometers from the station and missed milk samples. The licensee captured this issue in their CAP as IR 4002540. Licensee corrective actions included, but were not limited to, revising the applicable REMP procedures and investigating the possibility of growing the principal food pathways on the licensees owner controlled area or other approved licensee property within the 10 kilometer site radius. The performance deficiency was determined to be more than minor because it was associated with the Program and Process attribute of the Public Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of the public from radiation. Specifically, the licensee failed to implement effective sample collection from sample locations for food products from three of the major quadrants during annual REMP sampling and collections in 2015. The licensees Offsite Dose Calculation Manual (ODCM), as written, did not meet 10 CFR Part 50, Appendix I, which requires the licensee to establish and provide data on measurable levels of radiation and radioactive materials in the site environs. The finding was determined to be of very low safety significance in accordance with IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, because it only involved the licensees REMP. The inspectors determined that this finding had a 4 cross-cutting component in the area of human performance, change management aspect, because the licensee did not use a systematic process for evaluating and implementing changes in their REMP sampling and collection program. (H.3)
05000456/FIN-2017002-022017Q2BraidwoodFailure to Adequately Implement Technical Specification Surveillance Frequency Requirements into Implementing ProceduresGreen. A finding of very low safety significance and an associated NCV of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were identified by the inspectors for the licensees failure to have appropriate implementing procedures for TS SR 3.9.3.2. Specifically, procedure BwIS NR203, Post Accident Neutron Monitoring System Discriminator Adjustment, 3 did not provide for determining and checking the discriminator voltage for the system at an 18-month frequency, as specified by TS SR 3.9.3.2. The licensee entered this issue into their CAP as IR 4010147 with an action to revise the surveillance frequency to every 18 months for each channel. The performance deficiency was determined to be more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective. The finding screened as having very low safety significance (Green) because it did not result in the loss of operability or functionality of any SSC. The licensee performed a review of the records associated with the last three years of operation and did not find any instances in which the post-accident neutron monitors (PANMs) were used to satisfy TS 3.9.3, Nuclear Instrumentation, requirements. No cross-cutting aspect was associated with this finding because it was confirmed not to be reflective of current licensee performance due to the age of the performance deficiency.
05000456/FIN-2017002-012017Q2BraidwoodFailure to Adequately Implement Surveillance Frequency Program for the Deferral of a Technical Specification SurveillanceGreen. A finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.5.19.b, Surveillance Frequency Program, were identified by the inspectors for the licensees failure to implement the requirements contained in the surveillance frequency control program when making a change to the specified frequency of TS Surveillance Requirement (SR) 3.3.1.11. On May 3, 2017, the licensee improperly deferred a TS required surveillance through the preventive maintenance deferral process due to a belief that it was a preventive maintenance activity and not an activity supporting a TS SR. The licensee entered this issue into their corrective action program (CAP) as Issue Report (IR) 4009050 with an action to re-establish the surveillance at an 18-month frequency and to perform it before the end of the Unit 2 refueling outage (RFO) A2R19. The performance deficiency was determined to be more than minor because if left uncorrected it could lead to a more significant safety concern. The finding screened as being of very low safety significance (Green) because it did not result in the loss of operability or functionality of any system, structure, or component (SSC). The inspectors determined that this finding had a cross-cutting component in the area of human performance, work management aspect, because the licensee failed to utilize a work process that included proper coordination with different groups or job activities. Specifically, licensee personnel conducting the deferral did not coordinate the activity with personnel in either the operations or regulatory assurance departments. Knowledgeable personnel in either of these station organizations could have identified that the wrong process for deferral was being utilized. (H.5)
05000266/FIN-2017001-012017Q1Point BeachLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an NCV of TS 5.5.14, Safety Function Determination Program (SFDP), due to the failure to detect a loss of safety function and ensure appropriate actions were taken during maintenance activities conducted during performance of WO 40513133 for troubleshooting the check source drive mechanism for RE235, control room noble gas monitor, on January 18, 2017. In addition to the troubleshooting activities in WO 40513133, the licensee concurrently performed preventative maintenance on W14A, F16 control 32 room charcoal filter fan, and W13B2, control room recirculation fan. Due to these activities, the licensee implemented procedure NP 10.3.8, Safety Function Determination Program, to ensure that a loss of safety function was detected and the appropriate actions were taken for the equipment out of service associated with the CREFS. Specifically, NP 10.3.8, step 4.2.2 stated, Perform Loss of Safety Function Evaluation. Contrary to NP 10.3.8, step 4.2.2, an adequate loss of safety function evaluation was not performed for the CREFS system based on the equipment that was out of service. As a result of the inadequate loss of safety function evaluation, the licensee did not perform the Required Actions of TS limiting condition for operation (LCO) 3.7.9, Control Room Emergency Filtration System (CREFS), Condition C. The inadequate loss of safety function evaluation was identified when an operator wrote an action request that questioned condition of the CREFS during maintenance activities on January 18, 2017. TS 5.5.14, Safety Function Determination Program, required, in part, that if a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. Contrary to the above, on January 18, 2017, the licensee did not enter the appropriate Conditions and Required Actions of the LCO in which a loss of safety function existed. Specifically, the licensee did not adequately implement procedure NP 10.3.8, step 4.2.2, which resulted in the licensee not performing the Required Actions of TS LCO 3.7.9, Condition C. The licensee entered this issue into the CAP as AR 02183341. The inspectors determined that this issue was of very low safety significance (Green) after reviewing IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated October 7, 2016 and IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated July 1, 2012. The inspectors answered Yes to Question 1 in Exhibit 3, Section C, Control Room Auxiliary, Reactor, or Spent Fuel Pool Building. This resulted in the finding screening as Green.
05000315/FIN-2017001-012017Q1CookFailure to Brief Worker Entry to High Radiation Area Resulting in the Unplanned Dose Rate AlarmGreen . A finding of very -low safety significance and an associated NCV of Technical Specification 5.7.1.b was self -revealed for the failure to a make radiation worker aware of the radiation dose rate before entering a high radiation area. The failure to brief the worker resulted in an unplanned electronic dosimeter dose rate alarm. The worker immediately exited the area and reported the event to the radiation protection staff. The license e entered the event into their CAP as AR 2016 13827. The inspectors determined that the performance deficiency was more than minor i n 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. Specifically, worker entry into a high radiation area without an adequate briefing could lead to unintended dose. The inspectors also identified an example in IMC 0612, Appendix E, which is similar to the performance issue. Therefore, t he finding was determined to be of very -low safety significance in accordance with I MC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008. The violation was of very- low safety significance (Green) 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 4 component in the human performance area, H.4, in the area of teamwork and communication and coordination across organizational boundaries, specifically between radiation protection staff and the individual . This resulted i n the worker proceeding into areas that they were not briefed to enter which contained unknown dose rates .
05000316/FIN-2017001-022017Q1CookImproper Disconnect OperationGreen . A self -revealed finding and associated NCV occurred on January 10, 2017, when the licensee caused a loss of a qualified off-site circuit while opening a disconnect on the Unit 2 reserve feed transformer. Regulatory Guide 1.33 requires procedures for operating the onsite and offsite electrical distribution system; however the licensee did not develop a procedure or instruction for operating the electrical distribution system. Licensee personnel opened a disconnect to the Unit 2 reserve auxiliary transformer with the transformer energized but unloaded. This action resulted in trip of an upstream breaker and unplanned Technical Specification entry for the opposite unit. The licensee recovered the offsite circuit for Unit 1. The licensee entered the issue into the corrective action program (CAP) as Action Request ( AR ) 2017 0346. The inspectors determined that the failure to develop, implement , and maintain procedures or work instructions for the electrical distribution system was a performance deficiency. The performance deficiency impacted the mitigating system performance objective of ensuring the availability of systems that respond to initiating events. The finding was not greater than green in accordance with IMC 0609, Appendix A, Exhibit 2, dated June 19, 2012, because the answer to all four questions was no. The finding does not include a cross -cutting aspect because the licensee followed guidance for operating the disconnect that existed for the life of the plant and is therefore not reflective of current performance.
05000315/FIN-2017001-032017Q1CookFailure to Control Nonconforming Delivery Valve Holders on Emergency Diesel GeneratorsGreen . A self -revealed 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 XV, Nonconforming Materials, Parts, or Components, occurred when the delivery valve holder (DVH) on a fuel injection pump failed during a run of the 1AB emergency diesel generator ( EDG ). Each cylinder on an EDG has a fuel injection pump. The DVH is the part of the fuel injection pump where the high pressure fuel line meets the pump discharge. A thru- wall crack developed from a machined portion inside the DVH that had too sharp of a corner. This same phenomenon occurred onsite and caused a leak in 2013 as well. In 2013, the licensee identified the tight radius as an issue and also identified a particular manufacturing lot of DVHs that might have the tight radius. Contrary to their commercial grade dedication (CGD) procedures, the licensee did not 3 update their CGD plan for these parts to include the radius as a critical characteristic. Further, the licensee relied on informal communications from the commercial grade supplier of the parts to conclude only a certain subset of the suspected lot of DVHs were susceptible to cracking. Finally, several management -approved actions to remove all affected DVHs of the lot were not performed, as there was the belief by some that only certain DVHs were affected. As a result, the licensee installed many DVHs from the suspect lot they thought were acceptable. However, in December 2016, one of the DVHs thought to be acceptable developed a leak during an EDG run. The radius was discovered to be out of tolerance, as were numerous other radii in DVHs across all of the EDGs which were from the suspect manufacturing lot. The licensee declared three of the four onsite EDGs inoperable, replaced DVHs, and commenced a root cause evaluation to address the issue. The issue was more- than -minor because it adversely affected the Design Control attribute of the Mitigating Systems cornerstone. Specifically, allowing nonconforming parts to be installed on safety -related equipment without proper controls or evaluation adversely affects 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 Green because performance testing of representative DVHs and engine analysis demonstrated that the EDGs in the as -found condition would have been able to perform their safety functions for the required lengths of time. The inspectors determined the issue had a cross -cutting aspect in the Problem Identification and Resolution area, specifically, P.2, Evaluation. Despite identifying a defect on a safety related part due to a failure in 2013, the licensee failed to properly evaluate the condition and ensure all susceptible parts were accounted for. Specifically, the failure to follow station processes for corrective action and CGD resulted in a defective part causing a leak on an EDG
05000315/FIN-2017001-042017Q1CookLicensee-Identified ViolationIn LER 05000315 2015 004, the licensee identified multiple violations of TS 3.4.11, which requires each PORV and associated block valve to be operable. The licensee identified that on 3 occasions for Unit 1 and 4 for Unit 2 that one PORV was not operable when the supporting control air compressor was out of service. In these 7 cases, the licensee failed to close the associated block valve within 1 hour, as required by Condition B. Further, the licensee also failed to be in mode 3 within 6 hours and in some cases mode 4 as required by Condition H. The licensee failed to meet these requirements as follows: Unit 1: February 27, 2013 November 8, 2014 May 6, 2015 Unit 2: May 14, 2013 January 16, 2014 31 March 24, 2015 May 4, 2015 Th e inspectors evaluated the condition in accordance with IMC 0612, Appendix B and determined the issue was more than minor because it adversely affected the mitigating system cornerstone objective of ensuring the availability of systems that respond to initiating events. The inoperability impacted the equipment performance attribute of availability. Using IMC 0609, Appendix A , Exhibit 2, the inspectors answered all the questions no; therefore, the inspectors determined the finding was of very low safety significance. The licensee documented the issue in AR 2015 11204.
05000315/FIN-2017001-052017Q1CookLicensee-Identified ViolationThe inspectors reviewed AR 2017 0503, VT- 2 examination not completed. The AR documented several cases where the licensee identified that safety related equipment had been returned to service without the necessary American Society of Mechanical Engineers Code required exams or evaluations being done to satisfy post -maintenance test (PMT) requirements. TS 5.4, Procedures, requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in RG 1.33, Revision 2, Appendix A, February 1978. RG 1.33 Section 9 states, in part, that maintenance that can affect the performance of safety -related equipment should be properly preplanned and performed in accordance with written procedures or documented instructions appropriate to the circumstances. Contrary to this requirement, safety related valves 2 NCR 252 and 2 CMO 410 were returned to service approximately December 9, 2016 and November 13, 2016, respectfully, following maintenance without the required American Society of Mechanical Engineers visual inspections having been planned into the work orders. Once identified, the PMTs were verified complete or acceptable on January 24, 2017 for 2 NCR 252 and January 30, 2017 for 2CMO 410. The issue was more than minor because it adversely affected the Procedure Quality attribute of the Mitigating Systems cornerstone and was a programmatic issue. The finding screened as a Green NCV because operability was maintained as verified later via the appropriate PMTs . The licensee documented the issue in the aforementioned AR
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).
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) .
05000454/FIN-2017001-012017Q1ByronLicensee-Identified Violation

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

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

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-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).
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.
05000456/FIN-2016007-012016Q3BraidwoodIdentification of SCAQs in Accordance with the QATRThe team identified an Unresolved Item (URI) regarding the identification of significant conditions adverse to quality (SCAQs) in the CAP. Specifically, the team determined that the CAP, as implemented by PI-AA-125,Corrective Action Program, and PI-AA-120, Issue Identification and Resolution, appeared to not ensure that SCAQs were appropriately identified and corrected to prevent recurrence. Chapter 16 of the Braidwood Quality Assurance Topical Report (QATR) describes the licensees program to identify and correct conditions adverse to quality. Procedure PI-AA-125 implemented the requirements established in the QATR. During this inspection, the team reviewed the CAP procedure to determine how it ensured that SCAQs were identified and resolved. As part of this review, the team requested a copy of identified SCAQs over the last two years and were subsequently informed that none had been identified. Issue #1 - The team reviewed the QATR and noted that the following requirements applied: Section 2.1 stated that measures are required to assure that the cause of any significant condition adverse to quality is determined and that corrective actions to prevent recurrence (CAPRs) are implemented. Section 2.2.1, Significant Conditions Adverse to Quality, stated that in cases of significant conditions adverse to quality the cause of the condition must be determined and documented, the resolution determined and documented, and the corrective actions taken and documented to prevent recurrence. Step 2.116 of Appendix D of the QATR defined a significant condition adverse to quality as, a condition, which if left uncorrected, could have a serious effect on safety or operability. The team reviewed procedure PI-AA-125 and PI-AA-120, which delineated the process for the identification and screening of issues, and identified that these procedures did not include a provision to classify an identified issue as a SCAQ. The team also noted that the definition of a SCAQ was not being used to determine whether a RCE was needed; therefore, a CAPR did not appear to be directly associated with a SCAQ. Based on the above, the team questioned whether CAP procedure PI-AA-125 prescribed a process through which SCAQs were identified and documented, and corrective actions taken and documented to prevent recurrence as required by the QATR. The team discussed this issue with the licensee. The licensee stated that since the terms SCAQ and condition adverse to quality (CAQ) were not explicitly defined in NRC regulations, that they had created a graded approach of significance level and likelihood (which included risk and uncertainty) to ensure that items were properly dispositioned and the level of resources and rigor applied appropriately followed the CAP governance. The licensee further stated that the graded approach, along with a well-trained management team that has nuclear safety and conservative decision-making as their primary focus, provided for an effective CAP. Finally, the licensee stated that even if a CAPR was not issued, that CAs would prevent recurrence of the events entered into the CAP. The team questioned whether a CAPR and a CA would be equally effective as corrective actions to prevent the recurrence of issues dispositioned in the CAP. The licensee agreed that the two types of CAs were treated differently. For example, 1) the MRC was required to assess changes to the intent of a CAPR, which was not required for a CA, 2) an effectiveness review may not necessarily be assigned if an issue was corrected using only a CA, and 3) if there was a desire to suspend or modify a previously implemented CAPR, then a risk analysis and MRC concurrence was necessary; which was not the case for a CA. At the end of the inspection it was not clear how procedures PI-AA-120 and PI-AA-125 ensured that SCAQs were identified and documented, and corrective actions taken and documented to prevent recurrence. Additionally, it was not clear if the licensees process implemented the requirements in the QATR. Resolution of this issue will be based on additional NRC review to determine if a violation of NRC requirements occurred. Issue #2 - The team identified an example of a potential SCAQ for which the licensee implemented CAs that failed to prevent the issue from recurring. Specifically, for a December 30, 2013 oil leak on the inboard bearing housing of the Unit 1 Train B (1B) SX pump, the licensees CAs restored operability, but were not adequate to prevent recurrence and consequently an oil leak recurred on November 18, 2014. Both of these oil leaks resulted in the licensee declaring the 1B SX pump inoperable and required entry into Technical Specification (TS) Limited Condition for Operation (LCO) 3.7.8 (reference Non-Cited Violation (NCV) 05000456/201400502; Failure to Correct Undersized Essential Service Water Pump Bearing Casing Drain Line Resulted in System Inoperability). The team questioned whether the oil leaks on the inboard pump bearing housing of the 1B SX pump should have been categorized as a SCAQ as defined in the licensees QATR. Specifically, QATR Section 2.116, Definitions, defined a SCAQ as, A condition, which if left uncorrected, could have a serious effect on safety or operability. In this case, although the oil leakage at the inboard pump bearing housing first identified in 2013 was specifically addressed through repairs, the CAs were not adequate to prevent recurrence and a second oil leak occurred in 2014 that caused a serious effect on the operability of the 1B SX pump (i.e. rendered the 1B SX pump inoperable). Additionally, the team considered this issue to have a potentially serious effect on operability, because if left uncorrected the oil leakage would have depleted the oil supply reservoir resulting in a loss of lubrication to the pump shaft bearings that could damage the pump shaft and require substantial repairs to return the pump to operation. The team discussed this issue with the licensee. The licensees response was that because there was no potential for common cause failure, and there was no significant change to plant risk after removing the 1B SX pump from service, the events discussed above were appropriately screened as Significance Level 3 issues. The licensee also stated that a SCAQ would typically be assigned for a Significance Level 1 or 2 issue, but even if an issue was assigned this level of significance, it would not necessarily be categorized as a SCAQ. At the end of the inspection it was not clear how the definition of SCAQ in the QATR was utilized in the CAP. Resolution of this issue will be based upon additional NRC review and a determination of whether the failure of the 1B SX pump constituted a SCAQ as defined in the QATR.
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