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05000263/FIN-2018012-032018Q3MonticelloLicensee-Identified ViolationThis violation of very-low safety significance was identified by the licensee and has been entered into the licensee CAP. Therefore, this finding being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.Enforcement:Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Updated Final Safety Analysis Report, Appendix I,Evaluation of High Energy Line Breaks Outside Containment,Table I.5-2, Table of System Effects,Revision 36P, listed the Division II emergency power system as available during HELBs outside containment. Contrary to the above, on July 29, 1974, the licensee failed to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically,the Division II emergency power system would not be available during a HELB outside containment.Procedure B.09.07-05, Operations Manual Section 4.16 kV Station Auxiliary, Revision 53,had actions that required entry into the lower 4kV area to permit repowering Division II emergency power systems but this area would be inaccessible during the event. Significance: The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences.Specifically, the performance deficiency resulted in a condition were the Division II emergency power system would not be available during HELBs outside containment. The inspectors assessed the significance of the finding using the SDP in accordance with IMC 0609, 11 Appendix A, The Significance Determination Process for Findings At-Power, using Exhibit 2, Mitigating System Screening Questions,and concluded the violation was of very-low safety or security significance (Green)because the licensee reasonably demonstrated an alternate strategy was available to timely reach and maintain cold shutdown conditions. Corrective Action References: CAP501000011837, CAP 50100001593
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
05000263/FIN-2018012-022018Q3MonticelloFailure to Implement Adequate Freeze Protection Monitoring for Condensate Storage Tank Instrumentation Piping in Response to Industry Operating ExperienceThe inspectors identified a Green finding and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to establish measures to ensure conditions adverse to quality are promptly identified and corrected. Specifically, the licensee failed to identify that monitoring of the CST instrument line heat tracing performed every 30 days was inadequate to assure the safety-related CST level instrumentation remained operable during extreme cold weather conditions
05000263/FIN-2018012-012018Q3MonticelloInboard Main Steam Isolation Valve Closure Time Test Acceptance Criteria Did Not Account for the Design Basis Accident Containment Back Pressure and Pneumatic Supply Operating PressureThe inspectors identified a Green finding and an associated NCV of Title 10 of the Code of Federal Regulations(CFR), Part 50, Appendix B, Criterion XI, Test Control, for the failure to assure that applicable requirements and acceptance limits contained in the inboard main steam isolation valve (MSIV) design documents were incorporated into their test procedure. Specifically, the inboard MSIV closure time acceptance criteria contained in Functional Test Procedure 0255-07-IA-2, Main Steam Isolation Valve Functional Checks Test, did not account for the elevated containment pressure and the expected lower pneumatic supply pressure expected during design basis accidents.
05000255/FIN-2018001-032018Q1PalisadesLicensee-Identified ViolationA violation of very low safety significance (Green) was identified by the licensee, has been entered into the licensees corrective action program, and is being treated as a Non-Cited Violation consistent with Section 2.3.2 of the Enforcement Policy. Enforcement:Violation: Technical Specification 3.7.6 requires that the combined useable volume of the Condensate Storage Tank (CST) and Primary Makeup Storage Tank (T81) shall be greater or equal than 100,000 gallons. LCO 3.7.6, Condition A states that if the useable volume is not within this limit then A.1 Verify OPERABILITY of backup water supplies in 4 hours andA.2 Restore condensate volume to within limit in 7 days. Condition B states that if the Required Action and associated Completion Time is not met then B.1 Be in MODE 3 in 6 hours and B.2 Be in MODE 4 without reliance on steam generators for heat removal in 30 hours. Contrary to the above, on December 7, 2017 and March 3, 2016, the licensee failed to enter and comply with the actions required by LCO 3.7.6 Condition A and Condition B when Primary Makeup Tank Makeup Control Valve CV2008 could not be fully opened, resulting in a combined useable volume of the CST and T81 of less than 100,000 gallons.Significance/Severity Level: The inspectors answered No to all the questions in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, because even though the CST and T81 volume were considered inoperable by the TS requirements, there was not a loss of safety function because credited backup water sources were available and operable.Therefore, the finding screened as Green.Corrective Action References: The licensee entered these issues into their CAP as CRPLP20175589, CRPLP20175554, CRPLP20175551, and CRPLP20161116
05000255/FIN-2018001-022018Q1PalisadesLicensee Implementation of Enforcement Guidance Memorandum 15002, Enforcement Discretion for Tornado-Generated Missile Protection NoncomplianceOn June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (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 noncompliances 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, or 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 Palisades, the EGM provided for enforcement discretion of up to 3 years from the original date of issuance of the EGM. On December 7, 2017, and as supplemented on January 18, 2018, Palisades submitted a request to the NRC to extend the enforcement discretion from June 10, 2018 to June 10, 2020 (ML17341A415 and ML18018A328, respectively). By letter dated February 16, 2018, the NRC granted the request to extend enforcement discretion until June 10, 2020 (ML18046A675). The EGM permitted 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 provide 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 about 60 days of issue discovery. In accordance with the EGM, the comprehensive compensatory measures are toremain 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. Palisades was licensed prior to issuance of Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants (GDC). Specifically, GDC 2, Design Bases for Protection Against Natural Phenomena, and GDC 4, Environmental and Dynamic Effects Design Basis, discuss how SSCs important to safety shall be designed to protect against natural phenomena, such as tornadoes and shall be adequately protected against the dynamic effects of tornadoes, including protection against missiles. Palisades site-specific licensing bases compliance with GDC 2 and GDC 4 are described in the Updated Final Safety Analysis Report (UFSAR) Sections 5.1.2.2 and 5.1.2.4. Palisades protection of SSCs against tornado-generated missiles is also discussed in UFSAR Section 5.5, Missile Protection. On January 31, 2018, the licensee initiated condition report (CR) CRPLP201800556, which identified a nonconforming condition in the Palisades licensing basis. Specifically, the surge line from the component cooling water (CCW) surge tank to the CCW suction line was identified to be potentially vulnerable to a tornado missile through a doorway. The licensee previously identified a CCW system-related vulnerability on March 29, 2017. The March 29, 2017 CCW vulnerability and five additional vulnerabilities of other SSCs, which all received enforcement discretion, are documented in NRC Inspection Report 05000255/2017002 (ML17220A349). The licensee assessed this new vulnerability and concluded that previously established compensatory measures for the CCW system were adequate and no additional comprehensive compensatory actions were required. Therefore, the licensee declared the SSC operable, but nonconforming because no additional compensatory measures designed to reduce the likelihood of tornado-generated missile effects were required and the previously implemented compensatory measures were still in place. Corrective Action: The licensee documented the condition of the SSC in the CAP and documented the SSC as operable but nonconforming.Corrective Action Reference: CRPLP201800556 Enforcement: Violation: Enforcement discretion was applied to the required shutdown actions of the following Technical Specification (TS) Limiting Conditions for Operation (LCOs): TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs); andTS 3.7.7, Component Cooling Water (CCW) System.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 EGM 15002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Non-Compliance (ML16355A286). 11 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 implemented more comprehensive compensatory actions to resolve the nonconforming conditions within the required 60 days. These comprehensive measures were to remain in place until permanent repairs were completed, which for Palisades were required to be completed by June 10, 2020, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC such that discretion was no longer needed.The disposition of this enforcement discretion closes LER 05000255/201700101, Inadequate Protection from Tornado Missiles Identified Due to Nonconforming Design Conditions.
05000255/FIN-2018001-012018Q1PalisadesFailure to Maintain an Appropriate Documented Work Instruction for Reassembly of Primary Makeup Tank Makeup Control Valve CV2008A self-revealed Green finding and an associated NCV of Technical Specification 5.4.1, Procedures, was identified for the licensees failure to have an adequate maintenance work instruction for the reassembly of Primary Makeup Tank Makeup Control Valve CV2008. Specifically, because a previous CV2008 maintenance activity failed to properly set the height of the CV2008 jam nuts, the valve guide key fell out of place and in December 2017, CV2008 was unable to be manually stroked during surveillance testing
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.
05000315/FIN-2017004-042017Q4CookFailure to Verify the Adequacy of the Design for a Temporary ModificationA finding and associated violation of 10 CFR 50 Appendix B Criterion III self-revealed when licensee personnel could not obtain a water sample from a location designated as a connection point for a safety related temporary modification. Specifically, the licensee developed a temporary modification to add water to CCW but failed to verify the adequacy of the design in that the licensee did not validate the connection point could supply sufficient water as a source for CCW make-up. As an immediate action the licensee reestablished flow through the valves. The inspectors determined that the licensees failure to verify the adequacy of the design for the temporary modification was more than minor because it was associated with equipment performance attribute of Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance (Green) because the finding affected the qualification of CCW but did not render it inoperable. In this case, CCW remained operable based on credit taken for isolation valve capability. The finding includes a cross-cutting aspect in the human performance area of H.14, Conservative Bias.
05000315/FIN-2017004-032017Q4CookFailure to Promptly Correct The CAQ by Not Testing the CCW Leak Isolation ValvesThe inspectors identified a finding and associated NCV of Title 10 of the Code of Federal Regulations (CFR) Title 50, Appendix B Criterion XVI for failing to promptly correct a condition adverse to quality (CAQ). Specifically, in Inspection Report (IR) 05000315/3162015008 the NRC issued an NCV of 10 CFR 50 Appendix B Criterion III for the licensees failure to leak test isolation valves between redundant trains of the component cooling water (CCW) systems for Units 1 and 2. Despite opportunities to restore compliance, for Unit 1, the licensee suffered the violation from November 17, 2015, through November 4, 2017. As of December 31, 2017, the licensee continues to be in violation on Unit 2. The licensee tested the Unit 1 isolation valves during the fall 2017 outage and has scheduled testing of the Unit 2 valves in the spring 2018 outage. The inspectors determined that the licensees failure to promptly correct the CAQ by not testing the CCW leak isolation valves or otherwise restoring compliance was more than minor. The inspectors determined the issue was more than minor because it adversely affected the Mitigating Systems cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The issue was not greater than green because it did not render CCW inoperable. The inspectors determined the finding included a cross-cutting aspect of H.1, Resources.
05000315/FIN-2017004-022017Q4CookUnit 1 Letdown System Safety Valve Lift During Preparations for CooldownRefueling Outage Activities a. Inspection Scope The inspectors reviewed the Outage Safety Plan (OSP) and contingency plans for the Unit 1 refueling outage (RFO), conducted September 13 through November 26, to confirm that the licensee had appropriately considered risk, industry experience, and previous site-specific problems in developing and implementing a plan that assured maintenance of defense-in-depth. During the RFO, the inspectors observed portions of the shutdown and cooldown processes and monitored licensee controls over the outage activities listed below: licensee configuration management, including maintenance of defense-in-depth commensurate with the OSP for key safety functions and compliance with the applicable TS when taking equipment out of service; implementation of clearance activities and confirmation that tags were properly hung and equipment appropriately configured to safely support the work or testing; installation and configuration of reactor coolant pressure, level, and temperature instruments to provide accurate indication, accounting for instrument error; controls over the status and configuration of electrical systems to ensure that TS and OSP requirements were met, and controls over switchyard activities; monitoring of decay heat removal processes, systems, and components; controls to ensure that outage work was not impacting the ability of the operators to operate the spent fuel pool cooling system; reactor water inventory controls including flow paths, configurations, and alternative means for inventory addition, and controls to prevent inventory loss; controls over activities that could affect reactivity; maintenance of secondary containment as required by TS; licensee fatigue management, as required by 10 CFR 26, Subpart I; refueling activities, including fuel handling and reactor assembly/disassembly; startup and ascension to full power operation, tracking of startup prerequisites, walkdown of the containment to verify that debris had not been left which could block emergency core cooling system suction strainers, and reactor physics testing; and licensee identification and resolution of problems related to RFO activities. Documents reviewed are listed in the Attachment to this report. Inspections activities performed in the third quarter coupled with those in the fourth quarter constituted one RFO sample as defined in IP 71111.2005. b. Findings (Opened) Unresolved Item 05000315/201700402, Unit 1 Letdown System Safety Valve Lift During Preparations for Cooldown Introduction: Shortly after the shutdown for the Unit 1 refueling outage in September 2017, the licensee was establishing conditions in the charging and letdown system for the upcoming cooldown. After lowering letdown flow and attempting to adjust pressure, a letdown safety valve lifted and failed to completely reseat. Review of plant parameters following the event revealed that the evolution created saturation conditions in the letdown system. Subsequently, the steam bubbles collapsed causing a water hammer that lifted and damaged a relief in the system. The event was discussed in Section 4OA3 of Inspection Report 05000315/05000316/2017003. Description: The inspectors reviewed the licensees follow up of the issue in the CAP and spoke to personnel in the operations and maintenance departments. The licensee identified potential issues in the areas of procedure adequacy, operator performance, and equipment performance. However, the inspectors could not reconcile information on plant conditions with licensees statements regarding the cause. Because of ambiguity regarding the cause, the inspectors could not determine whether the corrective actions taken by the licensee were adequate. The licensee determined that an apparent cause evaluation need not be done therefore the inspectors reviewed available data, including plant computer data and a prior event from 2004. Since it is unclear what, if any, performance deficiency exists associated with this issue, the inspectors determined an unresolved item (URI) was necessary pending further follow up of the issue.Following the lifting of the safety valve, the licensee isolated letdown to stop the remaining leakage through the valve. The licensee then cycled the valve sufficiently enough for it to reseat so letdown could be restored and the cooldown continued. The safety valve was later discovered to be damaged from the event, so it was also repaired. Walkdowns were also conducted of the letdown piping to ensure no damage had occurred during the pressure transient. As part of their corrective actions, the licensee made some changes to the letdown procedure, recalibrated a letdown flow control valve, and developed actions to cover the event and lessons-learned in training. However, as stated above, the inspectors were unable to determine if these were sufficient to address the prevailing cause of the issue. The inspectors developed a series of questions for the licensee to explore more of the details behind the various potential issues. In order close the URI, the inspectors need to review the licensees response to questions provided and review available documentation of the event. (URI 05000315/201700402, Unit 1 Letdown System Safety Valve Lift During Preparations for Cooldown)
05000315/FIN-2017004-012017Q4CookFailure to Correct Numerous Anchor Darling Double Disc Gate Valve Non-ConformancesThe inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action for the licensees failure to correct a design non-conformance reported to the licensee through two related 10 CFR Part 21 reports. In March 2013, the licensee identified that 28 safety-related Anchor Darling double disc gate valves (ADDDGVs) may not have been assembled with an assumed amount of valve stem to wedge pre-torque before the stem was pinned into the wedge. The licensee had restored compliance to only one of these valves and had no plans to restore quality to the remaining 27 valves prior to the inspection. The licensee entered the inspectors conclusions into their corrective action program (CAP) as AR 201710399. At the end of this inspection the licensees plan was to restore compliance by either correcting the Part 21 issue or changing the design to accept the stem not having any pre-torque into the wedge.The performance deficiency was determined to be more than minor because if left uncorrected could become a more significant safety concern. Specifically, the failure to correct the design deficiencies could result in the valve pin breaking and consequential valve damage if the valves were operated at a high enough torque and/or thrust value(s). The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality of Mitigating Systems. Specifically, the licensee performed an operability determination which concluded that all 28 valve wedge pins had not sheared based upon the known historic operational history, pin material properties, and for using stem to wedge thread friction in some cases. The inspectors determined that this finding was not indicative of recent performance and therefore did not have a cross-cutting aspect assigned.
05000346/FIN-2017003-052017Q3Davis BesseUltrasonic Testing Records to Support Fuel Selection Were Not Being MaintainedA Severity Level IV NCV of 10 CFR 72.174, Quality Assurance Records, was identified by the NRC inspectors for the failure of the licensee as of June 22, 2017, to maintain sufficient records to furnish evidence of activities affecting quality. Specifically, the licensee failed to maintain ultrasonic testing (UT) records which were relied upon to demonstrate that the spent fuel selected for loading in calculation CNF062.02055, Revision 0, was correctly classified as intact. The licensee documented this issue in its CAP as CR201706976 and took timely corrective actions.The inspectors determined that the violation was of more than minor significance using IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues. Example 9a is applicable to this issue in that the licensee failed to maintain UT records for many fuel assemblies classified as intact for loading, and this failure to maintain records was not an isolated incident of one or two instances. The violation screened as a Severity Level IV NCV. Cross-cutting aspects are not assigned to traditional enforcement violations.
05000346/FIN-2017003-042017Q3Davis BesseFailure to Perform 10 CFR 50.59 EvaluationA finding of very low safety significance and an associated NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, was identified by the NRC inspectors for the licensees failure to maintain a record of a change from a method described in the USAR to another method. Specifically, the licensee failed to perform a written evaluation for the change to USAR defined load factors based on the design basis American Concrete Institute (ACI) 31863 Code to less conservative load factors based on the ACI 31871 Code. The licensee entered this issue into its CAP as CR201703025. Planned corrective action includes updating the USAR to reflect the changes to the Design Criteria Manual (DCM) for the load factors incorporated in the 1971 ACI 318 Code. 3 The inspectors determined that the licensees failure to perform a written evaluation for this change was a performance deficiency. The finding was determined to be more than minor because the inspectors could not conclude that the implemented change would not result in a departure from a method of evaluation described in the USAR) used in establishing the design bases and therefore not require a license amendment. Because the inspectors could conclude that the concrete structures designed using ACI 31871 load combinations would still have sufficient structural capacity to perform their design basis safety functions during a seismic event, the finding was determined to have very low safety significance corresponding to a Severity Level IV violation per Example 6.1.d.2 of the NRC Enforcement Policy. The inspectors did not identify a cross-cutting aspect associated with the finding because the finding was not representative of current licensee performance.
05000346/FIN-2017003-032017Q3Davis BesseFailure to Perform Adequate Evaluation of Cask CraneComponents and Crane Support StructureA finding of very low safety significance and an associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, was identified by the NRC inspectors for the failure of the licensees design control measures to provide for verifying or checking the adequacy of design of the Auxiliary Building spent fuel cask crane and crane support structure elements. Specifically, calculations involving the Auxiliary Building structure, crane runway rails, crane rail clips, and rail clip bolts had not been verified or checked to ensure the requirements of Updated Safety Analysis Report (USAR) Section 3.8.1.2 were included. The licensee documented these issues in its Corrective Action Program (CAP) as CR201705071, CR201707084, and CR201707114, and initiated actions to restore compliance.The performance deficiency was determined to be of more-than-minor significance because it was associated with the Barrier Integrity cornerstone attribute of Design Control and adversely affected the cornerstone objective of providing reasonable assurance that the physical design barriers, i.e. the Auxiliary Building, protect the public from radionuclide releases caused by accidents or events. The inspectors screened the finding through Inspection Manual Chapter (IMC) 0609, Appendix A, The Significance Determination Process for Findings At-Power, using Exhibit 3, Barrier Integrity Screening Questions. The finding screened as of very low safety significance because the finding only represented a potential degradation of the radiological barrier function provided by the Auxiliary Building. The inspectors identified a Human Performance, Design Margin (H.6) cross-cutting aspect associated with this finding. Specifically, the licensee failed to ensure the Auxiliary Building structure, cask crane runway rails, rail clips, and rail clip bolts reflected the intended design margins established based on the design and licensing basis.
05000346/FIN-2017003-022017Q3Davis BesseAuxiliary Feedwater Pump 1 Bearing FailureA URI was identified by the inspectors relating to the final determination of the cause of the AFP 1 turbine inboard bearing failure. On September 13, 2017, the licensee performed the scheduled quarterlysurveillance test on AFP 1. This test requires the pump to run loaded with full flow of water, whereas the monthly test runs the pump only lightly loaded with water being 10 pumped through a minimum recirculation line. Within three minutes after the full flow adjustments were completed, the AFP 1 turbine inboard bearing high temperature alarm (>220 oF) actuated. The licensee verified the alarm was valid and manually tripped the AFP 1 turbine approximately 30 minutes after the alarm was received. Oil samples indicated bearing damage. The licensee disassembled the AFP 1 turbine bearing and observed bearing failure.Initial evaluation of the bearing by the licensee revealed that the damage was due to insufficient lubrication caused by low oil level. The oil level at the time of failure was within the indicated acceptable band of the oil sight glass, however, indicated band was significantly larger than the vendor recommended 3/8 inch and not at the correct height.The oil level in the sump was too low to sufficiently wet the oil slinger ring. This condition was determined to have existed since the previous pump quarterly test on June 21, 2017. After that test, a technician removed an oil sample, but did not replenish the oil. The oil level indicated low to mid band, but within the (incorrectly marked) acceptable range on the sight glass at the time. The licensee entered this issue into their CAP as CRs 201709443, 201709817, 201709527, and 201709857. Because the licensee had yet to complete their investigation and analysis of the event by the end of this inspection period, the issue is being treated as a URI pending the inspectors review of the licensees completed root cause evaluation. (URI 05000346/201700302, Final Cause Determination of Auxiliary Feedwater Turbine Bearing Failure)
05000346/FIN-2017003-012017Q3Davis BessePinched Wiring Causing the Failure of Fuses Y210 and Y214An unresolved item (URI) was identified by the inspectors relating to the significance of pinched wires and licensees understanding of the condition and the extent of cause and condition. On July 6, 2017, during a planned replacement of fuse Y204 in electrical cabinet Y2, unrelated fuse Y214 blew. Both fuses were scheduled for replacement as part of the licensees project to replace Shawmut A25X style fuses that are susceptible to premature failure. The failure of fuse Y214 was unexpected, and the licensee was not able to discern a direct cause. The licensee determined that the failure was the fuse itself being so unstable that any perturbation was enough to cause failure. This failure resulted in multiple systems being declared inoperable including AFP 2, safety features actuation system channel 2, decay heat removal system interlock, and radiation element RE8447. On August 8, 2017, the same electrical cabinet, Y2, was opened for replacement of fuse Y216. Following the replacement, fuses Y210 and Y214 blew. The licensee attempted replacement of the fuses, but the replacement fuses blew again, shortly after being repowered. Initial licensee evaluation of the condition revealed thatthe wire bundle running along the hinge side of the cabinet door was unconstrained and two of the wires had become pinched between the door and cabinet frame, which damaged the wire insulation and allowed the wires to short circuit against the cabinet frame. The failure of Y210 and Y214 resulted in multiple systems being declared inoperable including AFP 2, safety features actuation system channel 2, decay heat removal system interlock, and emergency diesel generator 2. The licensee removed and replaced the damaged portion of the wires and used wire ties to constrain the wire bundle. The licensee entered this issue into their CAP as CRs 201707196 and 201708185. Because the licensee had yet to answer NRC inspector questions pertaining to the corrective actions and extent of condition by the end of this inspection period, the issue is being treated as a URI pending completion of the inspectors review. (URI 05000346/201700301, Examination of Extent of Cause and Condition of Pinched Wires in Electrical Cabinets)
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.
05000266/FIN-2016004-022016Q4Point BeachLicensee-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. The licensee identified a finding of very low safety significance (Green) and an NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, due to the failure to properly implement instructions in Work Order (WO) 40461957 for the replacement of the power range nuclear instrument (NI) 1N-43 gain potentiometer vernier dial. Specifically, step 5 of the WO stated, Replace the gain pot vernier with the preset spare. Prevent movement of the potentiometer shaft as much as possible. Contrary to the WO instructions, the technician performing the work believed it was necessary to dial the gain potentiometer to zero before replacing the dial and in doing so caused the 1N-43 NI high flux trip function to become inoperable. This was identified when the control room operators observed the indicated NI power reading for 1N-43 decrease to 82 percent and questioned the technician performing the work about the observed power change. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings required, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to the above, on December 2, 2016, the licensee did not accomplish activities affecting quality in accordance with the documented instructions. Specifically, the licensee did not follow step 5 of the work instructions in WO 40461957, causing the NI high flux trip function for 1N43 to become inoperable. The licensee entered this issue into the CAP as AR 02172378. 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 no to all questions in Exhibit 2, Section C, Reactivity Control Systems. This resulted in the finding screening as Green.
05000266/FIN-2016004-012016Q4Point BeachScaffolds Constructed Without Required Engineering ApprovalGreen: A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by inspectors for the licensees failure to follow step 4.1.3 (2) of procedure MAAA1001002, Scaffold Installation, Modification, and Removal Requests. Specifically, the licensee failed to obtain and document engineering approval for multiple scaffolds constructed in the cable spreading room that did not meet the separation criteria of Attachment 1 of MAAA1001002. The licensees short-term corrective actions included obtaining the appropriate engineering evaluations for the affected scaffolding and conducting a stand-down and information sharing with the scaffold builders to ensure they were aware of the importance of obtaining engineering approvals. The finding was determined to be more than minor because the finding, if left uncorrected, had the potential to become a more significant safety concern. Specifically, if the licensee continued to construct scaffolding without obtaining required engineering approvals, scaffolding could be constructed that was not seismically qualified and adversely affect the operability of surrounding structures, systems, and components (SSCs). The inspectors concluded this finding was associated with the Mitigating Systems cornerstone. 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, issued on October 7, 2016. Specifically, the inspectors used IMC 0609, Appendix A, SDP for Findings At-Power, issued June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions to screen the finding. The finding screened as of very low safety significance (Green) because the inspectors answered "No" to the screening questions. This finding has a cross-cutting aspect of Teamwork (H.4), in the area of Human Performance, for the failure of individuals and work groups to communicate and coordinate their activities across organizational boundaries to ensure nuclear safety is maintained. Specifically, the scaffold building team failed to communicate with the engineering organization to ensure the engineering evaluations were complete.
05000263/FIN-2016007-012016Q4MonticelloInadequate Procedure for Identification of Significant Conditions Adverse to QualityThe inspectors identified a finding of very low safety significance and non-cited violation of Title 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to prescribe a procedure appropriate to the circumstances with respect to the identification of a significant condition adverse to quality (SCAQ). Specifically, FPPAARP01, CAP Action Request Process, provided an overly restrictive definition of what constituted a SCAQ. Consequently, the failure to provide an adequate definition of a SCAQ could result in a failure to identify a SCAQ and therefore, failure to implement corrective actions that preclude repetitive failures of safety-related equipment. The licensee entered this issue into the CAP as action request (AR) 1536735. The inspectors determined that the licensees failure to prescribe a procedure appropriate to the circumstances under FPPAARP01 was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," because, if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Although, this issue could potentially affect each of the Reactor Safety Cornerstones, the inspectors elected to evaluate this issue under the Mitigating Systems Cornerstone because inspectors concluded it impacted the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage) more than the attributes of the other Cornerstones. The inspectors utilized IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding screened as very low safety significance (Green) since the inspectors answered No to each of the questions in Exhibit 2, Section A, Mitigating Systems Screening Questions. The inspectors determined that the performance characteristic of the finding that was the most significant causal factor of the performance deficiency was associated with the cross-cutting aspect of Problem Identification and Resolution, Self-Assessment, and involving the organization routinely conducting self-critical and objective assessments of its programs and practices. Specifically, the failure to identify the overly restrictive definition of SCAQ during previous audits of the CAP was caused by an insufficiently self-critical audit focus.
05000282/FIN-2016007-022016Q2Prairie IslandInadequate Operability DeterminationsA finding of very low safety significance with two examples and an associated non-cited violation of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to accomplish the requirements of procedure FPOPOL01, Operability/Functionality Determination, Revisions 14 and 15. Specifically, on two occasions, the licensee failed to properly evaluate potential operability concerns associated with the Unit 2 emergency diesel generator (EDG) day tanks and the Unit 2 train A cooling water (CL) system piping. The licensee entered the issues into the Corrective Action Program as Action Requests 1525842 and 1526070. The inspectors determined that the licensees failure to accomplish the requirements of procedure FPOPOL01, Operability/Functionality Determination, Revisions 14 and 15, to properly evaluate the operability issues associated with the Unit 2 EDG day tank fuel oil level and the Unit 2 CL system piping (both safety-related, mitigating systems) was a performance deficiency. The performance deficiency, with two examples, was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," it was associated with the Mitigating Systems Cornerstone attributes of Equipment Performance (for the Unit 2 EDGs) and Protection against External Factors (for the Unit 2 CL piping) and adversely affected the Cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events. The inspectors utilized IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding screened as very low safety significance (Green) since the inspectors answered Yes to Question 1 of Section A of Exhibit 2, Mitigating Systems Screening Questions. The inspectors concluded that this issue was cross-cutting in the area of Problem Identification and Resolution in the aspect of Evaluation. As defined in IMC 0310, Aspects Within the Cross-Cutting Areas, this aspect states, The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee had not thoroughly evaluated the operability issues associated with the Unit 2 EDG day tank levels and the Unit 2 CL piping structural integrity.
05000282/FIN-2016007-012016Q2Prairie IslandFailure to Ensure Breaker Main Contacts are Fully AlignedA finding of very low safety significance and associated non-cited violation of Technical Specification Section 5.4.1, Procedures, was identified by the inspectors for the licensees failure to ensure the 21 safeguards diesel exhaust fan main contact connectors were fully engaged and aligned as required per electrical maintenance procedures to ensure proper operation of the breaker. As part of their corrective actions, the licensee aligned and re-engaged the main contact connectors as necessary. In addition, the licensee ensured maintenance personnel were aware of the operating experience to prevent the same issue from occurring in the future. The violation was entered into the licensees corrective action program as Action Request 1525844. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone and the breaker failure led to the inoperability of the 21 safeguards diesel exhaust fan and impacted the availability of the 22 cooling water system diesel driven pump. This finding represented a loss of the 22 safeguards diesel cooling water pump function for longer than the Technical Specification allowed outage time of 7 days and therefore required a detailed risk evaluation. The regional senior reactor analyst performed a detailed risk evaluation of this finding using the Prairie Island Standardized Plant Analysis Risk Model revision 8.19 and determined the finding was of very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because it was not indicative of current performance.
05000263/FIN-2016001-012016Q1MonticelloFailure to Use Procedures While Performing Activities Affecting QualityAn NRC identified finding of very low safety significance (Green) and associated of 10 CFR 50, Appendix B, Criterion V; Instructions, Procedures, and Drawings, was identified on February 5, 2016, as a result of the licensees failure to use procedures while performing activities affecting quality. Specifically, the licensee failed to accomplish activities affecting quality in accordance with FP-G-DOC-03; Procedure and Work Instruction Use and Adherence, in that documented procedures were not used to install a conduit support on safety related Emergency Filtration Train (EFT) Division II conduits. Immediate corrective actions included removal of the support and entering the issue into the licensees Corrective Action Program (CAP) 1511349. The finding was determined to be more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the inspectors based this determination on the fact that performing activities affecting quality without using procedures has the potential to adversely affect the design/qualification of a Structure, System, and Component (SSC) or impact the operability or functionality of a system or component. The inspectors determined the finding to have very low safety significance (Green). The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, teamwork because of the licensees work group failures to communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained.
05000263/FIN-2015003-042015Q3MonticelloDrywell to Torus Vacuum Breaker Past OperabilityDuring the cycle preceding the 2015 refueling outage, two evaluations associated with torus to drywell vacuum breaker operation were developed due to issues identified in the first quarter 2014. These included: CAP 1417977, Failure of drywell-torus vacuum breaker to close, which identified an occasion of dual indication during Procedure 0143 procedure. A second occurrence was observed several days later and was documented in CAP 1418471, AO-2382A Torus-to-DW vacuum breaker closed indication anomaly. CAP 1420318, DW-Torus vacuum breaker work performed with inadequate PMT, identified the PMT following shaft sealing component (O-ring) replacement during the 2013 outage was not performed as planned. The licensee evaluations for these CAP conditions concluded the Drywell to Torus vacuum breakers were operable. However, neither evaluation specifically considered the effect of an interference between the vacuum breaker test lever and vacuum breaker test actuator stem. Since this specific mechanism was not addressed in these two evaluations, past operability of the torus to drywell vacuum breakers was questioned. As a result, the licensee established a past operability evaluation be conducted via CAPs 1479198 and 1478212. The licensee completed its past operability evaluation on June 26, 2015. After review, the inspectors conveyed a number of questions to the licensees engineering staff in regard to the past operability evaluation. Although the licensee provided responses for the majority of these questions during the remainder inspection quarter, the licensee had requested external input in regard to one of the inspectors questions. Specifically, inspectors questioned whether it was possible for the bottom of the lever arm to be at an elevation above the top of the actuator stem at valve disc full open and if so, could the valve test lever arm have come to rest on top of the actuator stem, potentially impacting the ability of the vacuum breaker valve to close. Upon the close of this inspection period, that input had not yet been finalized and made available to the inspectors. As a result, this issue was considered to be an unresolved item pending a review of the licensees response and past operability for CAPs 1479198 and 1478212, including and the licensee response to open inspector questions.
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