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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000255/FIN-2013002-01 | 2013Q1 | Palisades | Failure to Establish an Acceptable Component Cooling Water Heat Exchanger Final Test Frequency | The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control , for failure to establish testing to demonstrate the safety-related Component Cooling Water (CCW) heat exchangers would perform satisfactorily in service. Specifically, the licensee failed to demonstrate the heat exchangers fouling factors would remain acceptable to ensure adequate heat transfer capability prior to changing the inspection, cleaning, eddy current testing, and thermal performance testing frequency to 12 years. The licensee entered this issue into their CAP as CR-PLP-2012-05132 and CR-PLP-2013-00544 and implemented actions to revise the inspection, cleaning, testing, and maintenance frequencies to less than 5 years. The issue was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability reliability and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the inappropriate test frequency affected the licensees ability to ensure the CCW heat exchangers were available and capable to reliably perform as expected. The finding screened as of very low safety significance (Green) because the inadequate test program was not a design deficiency and did not result in a loss of system or component function. This finding has a cross-cutting aspect in the area of human performance, decision making because the licensee did not use conservative decision making and did not conduct effectiveness reviews of safety significant decisions to verify the validity of underlying assumptions, identify possible unintended consequences, or determine how to improve future decisions. Specifically, the licensee failed to use conservative decision-making or verify the validity of underlying assumptions when evaluating the effect that reducing the frequency of testing, inspection, cleaning, and maintenance would have on the CCW heat exchangers. |
| 05000255/FIN-2013002-02 | 2013Q1 | Palisades | Inadequate Work Instructions for Component Cooling Water Heat Exchanger | The inspectors identified a finding of very low safety significance (Green) with an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to properly plan and document work on the safety-related A CCW heat exchanger during a forced outage to repair leaks in the heat exchanger. Contrary to Criterion V and site implementing procedures EN-DC-115, Engineering Change Process, and EN-WM-105, Planning, the licensee did not ensure that appropriate quantitative or qualitative acceptance criteria for determining that important activities affecting quality were included in the work done to re-plug a population of leaking tubes in the heat exchanger. The licensee changed the work instructions to include the acceptance criteria after questioning by the inspectors. The licensee also interviewed workers to ensure the criteria had been utilized during earlier plug installation. The licensee entered the issue into their CAP as CR-PLP-2013-00773 and CR-PLP-2013-00969. The issue was determined to be greater-than-minor per IMC 0612, Appendix B, Issue Screening, because if left uncorrected, it could lead to a more significant safety concern. The inspectors decision was informed by examples 3j and 3k in IMC 0612, Appendix E, Examples of Minor Issues. The examples refer to an issue not being minor if significant programmatic deficiencies were identified with the issue that could lead to worse errors if left uncorrected. When the issue was first raised by the inspectors, only one of the two critical parameters was initially added to the revised work instructions. Further, two examples of inadequate documentation were identified. A basis for removing steps to check for leaks was not properly documented; and it was not clear from the completed work packages that the engineering acceptance criteria were met. Given these issues, the inspectors determined the threshold for a finding was met. The inspectors concluded the finding adversely impacted the Mitigating Systems Cornerstone objective and was of very low safety significance (Green) utilizing IMC 0609, Significance Determination Process. Specifically, utilizing Exhibit 2 of Appendix A, all questions in Section A were answered no. The finding had an associated cross-cutting aspect in the work control component of the human performance area. Specifically, the licensee did not coordinate work activities by incorporating actions to ensure interdepartmental alignments were made while planning and executing the work to assure plant and human performance. |
| 05000255/FIN-2013002-03 | 2013Q1 | Palisades | Damage to A AFW Pump Packing During Surveillance Run | A self-revealed finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion V was identified for the failure to conduct the A Auxiliary Feedwater (AFW) pump technical specification surveillance test in accordance with the prescribed in-service test procedure. Specifically, plant personnel conducting the surveillance test on the A AFW Pump adjusted packing when it was not required per the guidance in the procedure, which caused the pump packing to overheat and start smoking, resulting in unplanned inoperability of the pump. The licensee documented the issue in their corrective action program as CR-PLP-2013-01128 and completed an apparent cause evaluation. Planned corrective actions included revising the in-service test procedure. The finding was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Mitigating Systems Cornerstone attribute of human performance and adversely impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, a packing adjustment was made without being required by the procedure, causing the pump to overheat, which resulted in unplanned inoperability of the safety-related and risk significant A AFW pump. The finding had an associated cross-cutting aspect in the area of human performance related to the cross-cutting component of resources, in that the licensee ensures plant personnel have complete, accurate, and up-to-date design documentation, procedures, and work packages. In this finding, the fact that the A AFW pump has a unique packing design was not evident in the procedure being used and was not discussed during the pre-job briefs. |
| 05000255/FIN-2013002-04 | 2013Q1 | Palisades | Failure to Perform DAC-Hour Tracking | The inspectors identified a finding of very low safety significance and an associated NCV of TS 5.4.1. Specifically, the licensee failed to perform Derived Air Concentration (DAC)-Hour tracking for airborne transuranic radioactivity as required by a quality plant procedure, EN-RP-131, Air Sampling, resulting in untimely internal dose assessments for selected plant workers. The issue was entered in the licensees corrective action program as CR-PLP-2012-02683. The licensees immediate corrective actions included re-evaluating the use of site-specific work instructions. Long-term corrective actions included procedure changes and completing the required personnel dose assessments utilizing upper bounding radiological conditions. The finding is more than minor because it was associated with 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, not performing DAC-Hour tracking for airborne transuranic radioactivity affected the licensees ability to assess workers internal exposures in a timely manner and adversely impacted the licensees ability to monitor, control and limit workers radiation exposures (committed effective dose equivalent or internal dose). In accordance with IMC 0609 Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding had very low safety significance (Green) because the finding: (1) did not involve as-low-as-is-reasonably-achievable (ALARA) planning and controls; (2) did not involve a radiological overexposure; (3) there was not a substantial potential for an overexposure; and (4) there was no compromised ability to assess dose. The inspectors determined that the primary cause of this finding was related to a cross-cutting aspect in the area of human performance, resources component, such that the licensee maintains complete, accurate and up-to-date procedures and work packages. |
| 05000255/FIN-2013002-05 | 2013Q1 | Palisades | Failure to Take Corrective Action to Prevent Recurrence of CRDM Pressure Boundary Leakage | A self-revealing finding of very low safety significance (Green) with associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, and Technical Specification (TS) 3.4.13, Primary Coolant System (PCS) Operational Leakage, was identified for failure to take corrective actions to prevent recurrence of Control Rod Drive Mechanism (CRDM) cracking and leakage, a significant condition adverse to quality (SCAQ). Specifically, for Criterion XVI the licensee failed to include the internal CRDM housing weld build-up area within the scope of corrective actions taken for a 2001 CRDM through wall leak on CRDM-21, caused by transgranular stress corrosion cracking (TGSCC). Subsequently, a through wall leak recurred in the weld build-up area on CRDM-24 in 2012 due to TGSCC. As a result, the licensee operated with PCS pressure boundary leakage, which is not allowed by TS 3.4.13. Further, because the licensee was not aware that the leakage was PCS pressure boundary leakage, the licensee did not implement the associated TS action statement. The licensee replaced CRDM-24 upper housing and entered the issue into their corrective action program as CR-PLP-2013-01134. Additional corrective actions are described in NRC Inspection Report 05000255/2012012. The inspectors determined that this issue was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Initiating Events Cornerstone attribute of equipment performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability. Specifically, the licensee did not take adequate corrective actions to prevent recurrence of leakage in CRDM housings, which represents pressure boundary leakage. The inspectors determined this finding was of very low safety significance (Green) because the leak would not have exceeded the reactor coolant system leak rate for a small LOCA and could not have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function. Specifically, the slow rate of change for leakage for TGSCC in type 316 stainless steel will experience leakage rates well below a small break LOCA, which would be observed through the crack, alerting operators to take action to shut down the plant prior to experiencing a component rupture. The cause of this finding, non-conservative decision making, occurred over 10 years ago and is well outside of the nominal 3 year period in IMC 0612 for cross-cutting aspects. Therefore, this is not indicative of current performance, because no other opportunities to identify the issue occurred during the previous 3-year period. However more recently, the licensee exhibited non-conservative decision making with respect to addressing the potential for CRDM housing cracking and leakage during the recent root cause (Section 4OA2.4 (b.2) of this report), resulting in another finding. This cross-cutting aspect will be captured through the other finding. |
| 05000255/FIN-2013002-06 | 2013Q1 | Palisades | Failure to Adequately Address the Generic Implications of the Cracking Identified in CRDM-24 | The inspectors identified a finding of very low safety significance (Green) with an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, for the licensees failure to accomplish quality activities in accordance with the prescribed procedures. Specifically, the licensee failed to adequately evaluate and document the generic implications of the cause of the 2012 cracking identified in CRDM-24 in accordance with a quality procedure, Procedure, EN-LI-118, Root Cause Evaluation. This issue was entered into the licensees Corrective Action Program (CAP) under CR-PLP-2013-01500. Subsequently, the licensee decided to revise the inspection plan to add additional corrective actions to inspect a sample of welds No. 3 and No. 4 for TGSCC during the upcoming refueling outage. The inspectors determined that this issue was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, absent NRC identification, the licensee would not have completed further evaluations or inspections of CRDM housing welds, which could have resulted in additional CRDM housing failure and leakage by TGSCC. In accordance with Table 2, Cornerstones Affected by Degraded Condition or Programmatic Weakness, of IMC 609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the finding was associated with the Initiating Events Cornerstone because the failure of a CRDM housing is a Primary System LOCA initiator contributor. Using Exhibit 1, Initiating Events Screening Questions, in IMC 0609, Attachment A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined this finding was of very low safety significance because the leak would not exceed the reactor coolant system leak rate for a small LOCA and would not have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function. Specifically, the slow rate of change for leakage for TGSCC in type 316 stainless steel will experience leakage rates well below a small break LOCA, which would be observed through the crack, alerting operators to take action to shut down the plant prior to experiencing a component rupture. The inspectors determined that the primary cause of the failure to adequately consider welds No. 3 and No. 4 in the generic implications section of the root cause report (RCR) related to the decision making cross-cutting component in the human performance area because licensee staff did not use conservative assumptions in decision making. Specifically, the licensee did not use conservative assumptions when excluding welds No. 3 and No. 4 as being susceptible to TGSCC when there was not enough information to exclude them from consideration. |
| 05000255/FIN-2013002-07 | 2013Q1 | Palisades | Licensee-Identified Violation | One licensee identified finding of very low safety significance and an associated NCV was reviewed by the inspectors. Title 10 of the CFR, 71.5 (a), requires, in part, that each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the Department of Transportation Regulations in 49 CFR 107, 171 through 180, and 390 through 397, appropriate to the mode of transport. Title 49 CFR, 172.202, requires, in part, that descriptions of hazardous material on shipping papers include the identification number prescribed for the material and the proper shipping name prescribed for the material. Contrary to the above, on May 20, 2011, the licensee failed to use the proper shipping name and proper UN (United Nations) number when shipping radioactive resin in a Type B Cask. This issue was documented in the licensees CAP in CR-PLP-2011- 02887. Immediate corrective actions included notifying appropriate personnel and correcting the shipping records. The finding was assessed in accordance with IMC 0609, Attachment D, Public Radiation Safety Significance Determination Process, and determined to be of very low safety significance (Green). Specifically, the inspectors determined that the finding did not involve the Radioactive Effluent Release Program or the Radiological Environmental Monitoring Program. The finding did involve the transportation of radioactive material. However, no external radiation levels or surface contamination levels were exceeded, the finding did not involve the certificate of compliance, there was no failure to make notifications or provide emergency information, there was no breach of the package during transit, and there was no low-level burial ground non-conformance. |
| 05000255/FIN-2012005-03 | 2012Q4 | Palisades | Safety Injection Refueling Water Tank Evaluation of Corrosion | After reviewing CR-PLP-2012-4451, the inspectors were concerned the associated aging effects of the accumulated water were not properly managed because the condition of the affected annulus region was not evaluated by the licensee. Specifically, the accumulated water beneath the floor of the SIRW tank created an environment that could promote corrosion of the tank floor, and this condition does not appear to be formally evaluated. Also, the potential for corrosion could be exacerbated by concrete-aluminum interaction. The inspectors required more information to determine whether this issue constituted a finding of significance. This issue is considered an unresolved item pending further NRC review of the licensees actions. |
| 05000255/FIN-2012012-01 | 2012Q3 | Palisades | TS for PCS Pressure Boundary Leakage | On August 9, 2012, licensee management decided a plant shutdown was necessary to investigate the source of elevated PCS unidentified leakage. At the point of the shutdown on August 12, 2012, unidentified leakage was approximately 0.3 gpm and had been gradually increasing. The plant was required to be shut down per TSs when the unidentified leak rate exceeded 1.0 gpm. The NRC had been monitoring the increased trend in unidentified leakage since July 2012, when the plant was restarted from a forced outage, to ensure that the plant was taking action as the leakage increased. The operators completed a controlled reactor shutdown on August 12, 2012. Following containment entry, the cause of the rise in leakage was determined to be from a steam leak originating on CRDM-24 housing. The licensee reported that the leakage was PCS pressure boundary leakage. The TS for PCS pressure boundary leakage was 0 gpm. The licensee concluded that they operated the plant in a condition prohibited by TS 3.4.13, PCS Operational Leakage, and reported this condition to the NRC in accordance with 10 CFR 50.72. The NRC had previously granted enforcement discretion for situations where the licensee met all associated NRC regulations with regard to CRDM nozzle inspections and the violation was the result of equipment failure that could not have been reasonably avoided or detected (Enforcement discretion was granted under EA-02-011 in 2002 for the 2001 Palisades pressure boundary leak). Whether or not the licensee appropriately implemented their quality control program during the manufacture and installation of the CRDM-24 housing in 2001 or whether this failure was the result of an unavoidable equipment failure, is an unresolved item pending the review of the licensees root cause evaluation. It is not known, at this time, what caused the flaw which eventually propagated. (Unresolved Item (URI) 05000255/2012012-01, TS for PCS Pressure Boundary Leakage) |
| 05000255/FIN-2012012-02 | 2012Q3 | Palisades | Potential Inadequate Degradation Evaluation of CRDM Housings | On August 12, 2012, the licensee shutdown the plant to investigate the source of elevated PCS unidentified leakage. During a containment walk down post shutdown it was discovered the source of the leakage was from the housing of CRDM-24. The leak was classified as pressure boundary leakage. The licensee performed NDE as well as destructive testing on the housing of CRDM-24 at a laboratory contracted by the licensee. The preliminary results from this laboratory testing concluded that cracking in the housing of CRDM-24 was attributed to TGSCC. This form of cracking is prevalent when susceptible material such as the housing of CRDM-24 (austenitic stainless steel) is subjected to a corrosive environment (some level of oxygen and a small amount of chlorides) under applied stresses. The corrosive environment in this case is the primary coolant and the applied stresses include thermal and structural stress. The preliminary results from this laboratory testing concluded that cracking in the housing of CRDM-24 was attributed to TGSCC. This form of cracking is prevalent when susceptible material such as the housing of CRDM-24 (austenitic stainless steel) is subjected to a corrosive environment (some level of oxygen and a small amount of chlorides) under applied stresses. The corrosive environment in this case is the primary coolant and the applied stresses include thermal and structural stress. The licensee also examined the fracture surface of the housing of CRDM-24 at the location of the through-wall crack and identified six concentric rings (beach marks) propagating in a radial direction from the inside diameter out towards the outside diameter of the housing. The licensee considered the width of each of these beach marks as indicative of consistent periods/intervals corresponding to crack growth during operation (corrosive environment and stresses present). The relatively narrow band between successive concentric rings was interpreted by the licensee to represent periods of crack arrest during refueling outages (housing not in contact with corrosive environment and stresses are relaxed). However, more oxygenated water is introduced when the coolant system is refilled after refueling, starting the crack propagation process again when stress is applied. Based on the above interpretations of the fracture surface of the housing of CRDM-24, the licensee extrapolated the crack growth rate for the through-wall crack that caused leakage in the housing. The licensee applied this calculated crack growth rate to consider TGSCC in the remainder of the 44 CRDM housings based on the maximum size crack that could avoid detection. This licensee assessment revealed a time frame of over four years for the crack to propagate through-wall. The results of this crack growth rate as applied to the remainder of the 44 CRDM housings was documented in the licensees technical justification for startup. The inspectors have outstanding questions and concerns with regards to the methodology used by the licensee to extrapolate a crack growth rate due to TGSCC from the fracture surface of the housing of CRDM-24. The NRC Office of Nuclear Reactor Regulation (NRR) has been involved with detailed technical discussions with the licensee regarding these questions and concerns over the calculated crack growth rates. It is also possible that the concentric rings represent different periods where the temperature and pressure stresses were removed during heatup and cooldown cycles, irrespective of reintroducing oxygen into the system. In this case, the timeframe would be around 2 years for crack propagation. The NRC concluded that it was still acceptable for startup, but future detailed inspections may be needed sooner than the licensees assessment. The resolution of this URI is pending further discussions and potential independent examination of the fractured housing of CRDM-24 by NRC staff. (URI 05000255/2012012-02, Potential Inadequate Degradation Evaluation of CRDM Housings) |
| 05000255/FIN-2012012-03 | 2012Q3 | Palisades | Potential Failure to Prevent Recurrence of a Significant Condition Adverse to Quality | On August 12, 2012, the licensee shutdown the reactor to investigate the source of elevated PCS unidentified leakage. During a containment walk down post shutdown it was discovered the source of the leakage was the housing on CRDM-24. The leak was classified as pressure boundary leakage. An event similar to this occurred in 2001 when the licensee discovered a steam leak in the housing of CRDM-21 which was also classified as pressure boundary leakage. Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion XVI, states, In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. This 2001 pressure boundary leakage event was classified by the licensee as a significant condition adverse to quality. Therefore, as part of the developed corrective action, the need to prevent recurrence was specified. The root cause of the 2001 event was determined to be TGSCC due to susceptible material, fabrication flaws and inherent stresses in the weld due to the design of the component. The corrective actions to prevent recurrence established by the licensee included making changes to the design, fabrication, and material of this component. Specific attention was paid to the pressure retaining welds. The location of the current leak was not in proximity to the pressure retaining welds; therefore, susceptibility in this area was not considered. The root cause evaluation for the current failure was ongoing at the end of the inspection; therefore, it cannot be definitively concluded that the failure mechanism for the current leak is the same as the cause identified in 2001 and the corrective actions to prevent recurrence were inadequate. Although the method of crack propagation was the same, TGSCC, the initiation mechanism for the crack is not yet known. There is no current safety issue as the housing was replaced and the current crack propagation assessments and extent of condition reviews performed provide reasonable assurance a through-wall crack will not occur. (URI 05000255/2012012-03, Potential Failure to Prevent Recurrence of a Significant Condition Adverse to Quality) |
| 05000255/FIN-2012003-01 | 2012Q2 | Palisades | Inadequate Design Margins for Evaluation of Leaking SIRWT Nozzles | The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion III, Design Control for the licensees failure to adequately evaluate leaking Safety Injection and Refueling Water Tank (SIRWT) nozzles during the application of American Society of Mechanical Engineers (ASME) Code Case N-705. During the April-May 2012 refueling outage, the SIRWT was drained for inspection and repairs and a deformed nozzle was sealed off, as it was believed to be the potential source of pre-outage leakage. Upon refill, leakage was observed under a different section of the roof upon which the SIRWT rests, indicating a potentially new leak. The licensee employed ASME Code Case N-705 to demonstrate tank operability given the existing leakage and set an upper limit for allowed leakage. Inspector review of the approved evaluation identified certain Code Case criteria that were not discussed, namely, the residual weld stresses and seismic sloshing stresses. After discussions with the inspectors, the licensee developed residual weld stress values for their evaluation and discussed potential effects of seismic sloshing. The result was a reduction in allowed leakage from 130 gallons per day (gpd) to 34.8 gpd. The licensee entered the issue in their CAP as CR-PLP-2012-04245 and CR-PLP-2012-03732. The finding was determined to be more than minor because the finding, if left uncorrected, could become a more significant safety concern. The inspectors utilized examples 3j and 3k in IMC 0612, Appendix E, Examples of Minor Issues, to inform this determination. Omission of Code-Case-required parameters in the approved evaluation led to reasonable doubt on the operability of the system had the licensee ascended to a mode requiring SIRWT operability. Further analysis was also required by the licensee. Absent NRC identification, the failure to adequately evaluate the leaking SIRWT nozzles could have allowed unstable cracks to remain in service. Unstable nozzle cracks could propagate and allow unacceptable leakage from the SIRWT resulting in loss of inventory and increase the risk for insufficient core cooling for post LOCA conditions. This finding impacted the Mitigating Systems Cornerstone attribute of Equipment Performance (reliability). The finding adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because the licensee promptly corrected this issue and lowered the amount of allowed leakage, the inspectors answered No to all of the worksheet questions identified in IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a for the Mitigating Systems Cornerstone. The correct leakage limit was in place prior to the required time the tank needed to be operable. Therefore, this finding screened as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance for the work practices component. The licensee did not provide adequate supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported (H.4.c). Specifically, the licensee failed to ensure that the vendor evaluation to demonstrate SIRWT nozzle integrity with through-wall cracks included consideration of residual weld stresses and seismic sloshing stresses. The inspectors determined the primary cause of this finding based upon discussions with the licensees engineering staff. |
| 05000255/FIN-2012003-02 | 2012Q2 | Palisades | Operation of Primary Coolant Pumps Outside Design Basis | The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 50 Appendix B, Criterion III, Design Control, for the failure to operate the Primary Coolant Pumps (PCPs) in accordance with their design operating criteria. In October 2011, a slight rise in vibration levels on the C PCP occurred and was sustained for approximately 24 hours. This was followed by a short spike in vibrations and a return to a lower stabilized value than what had been previously observed. Investigation by the licensee revealed it was likely a piece of an impeller vane which had deformed and broken free. Based on a review of operating experience associated with impellers and further licensee investigation, the inspectors concluded that the PCPs had been operated outside of their license/design basis as stated in the Updated Final Safety Analysis Report (UFSAR) with regard to minimum net positive suction head and maximum flow. Further, based on impeller-like pieces found in the reactor vessel in 2007 (which an apparent cause stated likely came from a PCP), and an operating history which indicated past occurrences of vane breakage and degradation, the inspectors concluded the licensee had the ability to foresee and correct the condition affecting the PCPs prior to the release of a piece in October 2011. The licensee entered the issue in their Corrective Action Program (CAP) as CR-PLP-2011-5744 and performed additional research into the phenomena leading to the impeller degradation. The PCP operating sequence was changed, an Operational Decision Making Issue was implemented, and efforts to explore further procedural changes are on-going to mitigate degradation of the impellers. The issue was determined to be more than minor because it impacted the Design Control attribute of the Initiating Events Cornerstone, adversely affecting the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the potential release of impeller pieces in the primary coolant system (PCS) challenges the cornerstone objective. The issue screened as Green, or very low safety significance, based on answering no to the Loss-of-coolant Accident (LOCA) initiator question under the Initiating Events cornerstone in IMC 0609, Attachment 4, Table 4a. This was based on a review of the licensees assessment by the regional inspectors, experts at the Office of Nuclear Reactor Regulation (NRR) and Office of Research in determining the deficiency would not likely be an impact to the coolant pressure boundary. The inspectors determined there was no associated cross-cutting aspect because the finding was not indicative of current licensee performance. |
| 05000255/FIN-2012003-03 | 2012Q2 | Palisades | Failure to Follow Work Management Process for Reactor Head Work | The inspectors identified a finding of very low safety significance with an associated NCV of Technical Specification (TS) 5.4.1, Procedures, for the failure to properly follow the work management process for work done to loosen stuck reactor head studs. During the April-May 2012 refueling outage, difficulty was encountered in loosening some of the reactor head studs to support refueling operations. The decision was made to retension the studs that had already been detensioned (without ascending back to Mode 5 from Mode 6) and start over using a more precise electric pumping unit that had not been used to that point due to equipment issues. Contrary to EN-WM-102, Work Implementation and Closeout, the licensee used the field change process, not authorized for this type of change, to pen-and-ink different tensioning values and sequence in the normal tensioning procedure (so as not to return to Mode 5). Additionally, the inspectors identified that the steps documented as having been performed as a record of the contingency actions taken differed from what was actually performed. The licensee entered the issue into the CAP as Condition Reports CR-PLP-2012-2610 and CR-PLP-2012-2848, and corrected the contingency work instructions. The issue was determined to be more than minor because if left uncorrected, it could lead to more significant safety issues. Specifically, the failure to follow appropriate processes and correctly document reactor head work is indicative of shortfalls that could occur for other safety-related work. Additionally, the licensee was slow to recognize the issue. The inspectors concluded that the Initiating Events Cornerstone was impacted because of the potential for an inadvertent mode change. The finding screened as Green, or very low safety significance, using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process, based on all of the mitigation criteria being met and no phase 2 or 3 analysis being required per Checklist 3, indicating there was no impact to shutdown safety functions. The inspectors determined that the finding had an associated cross-cutting aspect in the area of human performance in that personnel work practices did not support human performance. Specifically, supervisory and management oversight failed to assure the proper processes were followed |
| 05000255/FIN-2010004-06 | 2010Q3 | Palisades | Use of TLDs May Not Be Consistent With the Methods Used by the NVLAP Accreditation Process | The inspectors identified that the licensees use of dosimeters (TLDs) may not be consistent with the methods used by the NVLAP accreditation process. As a result, the inspectors identified an Unresolved Item (URI) for the apparent non-compliance with 10 CFR 20.1501(c)(2) because the accreditation process for the types of radiation included in the NVLAP program may not approximate the types of radiations for which the individual wearing the dosimeter is monitored. The licensee uses a vendor to supply and process dosimeters used to measure radiation exposure for the monitored workers. This vendor is NVLAP accredited for beta, gamma, neutron, mixture of beta/gamma, and mixture neutron/gamma radiations. However, the licensee uses the dosimeters when workers may be exposed to beta, gamma, and neutron radiations within the same monitoring period. The inspectors determined that this mixture of three radiation types may not be aligned with the accreditation process. The licensee concluded that since the processor was accredited for all three types of radiation, the dosimeter could be used when all three types are encountered during the same monitoring period. During the inspection, the licensee contacted the dosimeter processor for clarification. The licensee indicated that the dosimeter processor makes some assumptions of the types of radiation that the worker was exposed to and selects an appropriate algorithm for processing the dosimeter. These assumptions consider whether the reactor is creating power, therefore creating neutron radiation. The inspectors agreed that workers are not generally exposed to neutron radiation during periods when the reactor is not creating power. However, the licensees program does not require a dosimeter change when the reactor is shut down for maintenance or when the reactor is returned to power. This created a condition where the same dosimeter could be used in a manner that is not consistent with the methods described by the dosimeter processor. The issue remains under review by the NRC pending completion of the licensee\\\'s testing and evaluation of dosimeters exposed to all three types of radiation. The issue is categorized as a URI pending completion of that revised evaluation and the NRCs review of it (URI 05000255/2010004-06 Use of TLDs May Not Be Consistent With the Methods Used by the NVLAP Accreditation Process). |