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05000327/FIN-2017002-02Sequoyah2017Q2Licensee-Identified ViolationUnit 1 and Unit 2 facility technical specifications LCO 3.6.10 required two operable EGTS systems in Modes 1 through 4. Contrary to the above, on August 2, 2016,during a system review, plant engineers noted a design flaw that could have resulted in one train of EGTS being rendered inoperable since initial plant operation. This problem was entered into the licensees CAP as CR 1198440 and CR 1200028. The TVA probabilistic risk assessment model does not consider the EGTS in core damage and large early release frequencies. The EGTS system is designed to maintain the shield building at a negative pressure and filter any leakage past the steel liner during a design basis event. With the EGTS inoperable, dose would still remain below 10 CFR 100 limits. The finding was screened using IMC 0609, Appendix A At Power Operation, and was determined to be of very low safety significance (Green). According to Exhibit 3, an issue related to degradation of the radiological barrier function of the reactor building is considered to be of very low safety significance.
05000346/FIN-2016001-03Davis Besse2016Q1Shield Building Emergency Ventilation System Operability with Watertight Door No. 108 Inadvertently Left OpenThe shield building EVS functions to collect and process potential leakage from the containment vessel to minimize environmental activity levels resulting from all sources of containment leakage following a design-basis accident. The EVS is required to maintain a negative pressure (a minimum of 14 inch water gauge), with respect to outside atmosphere, within the annular space between the shield building and the containment vessel and in the penetration rooms following an accident. In addition, it is required to provide a filtered exhaust path from the shield building annulus and the penetration and pump rooms following an accident. The EVS consists of two independent and redundant trains. Each train consists of a prefilter, a high efficiency particulate air filter, an activated charcoal adsorber section for removal of gaseous activity (principally iodines), and a fan. Ductwork, valves or dampers, and instrumentation also form part of the system. The EVS boundary, consisting of various walls and doors within the plants auxiliary building, must be intact and functional to ensure EVS operability. Door No. 108, Emergency Core Cooling System Pump Room No. 115 to Detergent Waste Drain Tank to Clean Waste Receiver Tank, is one such plant door. At approximately 7:53 p.m. on March 21, 2016, with the unit in Mode 1 and operating at power, operations personnel discovered a plant watertight door, Door No. 108, open and unattended. The operations personnel immediately secured the door and informed operations on-watch management of the issue. The on-watch operations shift manager determined that because the door was fully functional and closed when he was informed of the issue that neither the door nor the shield building EVS was inoperable. He then contacted the licensees on-duty management team to discuss the issue. Collectively, the licensees personnel concurred with the shift managers operability decision and determined that the issue was not immediately reportable under 10 CFR 50.72(b)(3)(v) as an Event or Condition that Could Have Prevented Fulfillment of a Safety Function, since no SSCs had ever been declared inoperable. Subsequently, licensee engineering personnel reviewing the issue determined that based on exiting plant calculations and the area of the door that it was highly improbable that the EVS would be able to have met its specified safety function with Door No. 108 open and unattended. The licensee entered this issue into their CAP as CR 201603694. An investigation by the licensee into the issue identified that the door had been inadvertently left open by contractor workforce personnel approximately five minutes before it was discovered open by operations personnel. During the next few days while conducting their routine review of the licensees CAP entries, the inspectors took note of this issue and questioned the licensee regarding their decision not to report the matter under 10 CFR 50.72(b)(3)(v). Licensee management subsequently decided to perform a special test of the EVS with Door No. 108 in the open position (under the administrative control of a designated individual) to empirically determine the capability of the EVS in this condition. The test was performed during the afternoon/evening hours on March 25, 2016. Preliminary results indicated that the EVS passed, albeit by only 0.08 seconds. Because the licensee had not yet completed their analysis of the issue following the March 25, 2016, special EVS test at the end of the inspection period, the issue is being treated as a URI pending the inspectors receipt and review of the licensees completed CAP documents and evaluation. (URI 05000346/201600103)
05000390/FIN-2015004-04Watts Bar2015Q4Shield Building Operability RequirementsThe inspectors identified an unresolved item (URI) associated with the requirements of Watts Bar Unit 1 technical specification (TS) 3.6.15, Shield Building. Additional inspection is required to determine if the requirements of 3.6.15.B applied during a specific testing alignment. On September 10, 2015, the licensee conducted 0-SI-65-6-A, Emergency Gas Treatment System (EGTS) Train A 10-Hour Operation. During the 10-hour time period of the test when the EGTS was in service, the auxiliary gas building treatment system was also in service for a Unit 2 construction test. This unique ventilation combination is not normally experienced during the 0-SI-65-6-A surveillance. As a result, shield building annulus differential pressure fell below the limit established by TS surveillance requirement (TSSR) 3.6.15.1 limits for the entire duration of the 10-hr EGTS surveillance. TS limiting condition for operation (LCO) 3.6.15.B requires annulus pressure be restored when it is outside of limits with a required completion time of 8-hrs. The licensee considered the note associated with TS LCO 3.6.15.B, which states that the annulus pressure requirement is not applicable during ventilating operations, required annulus entries, or auxiliary building isolations not exceeding one hour in duration. The licensee considered the alignment they were in at the time to be ventilating operations and thus the requirements of TS LCO 3.6.15.B did not apply. The licensee further considered that the note, as written, allowed grace from the annulus pressure requirement for ventilating operations for an unlimited amount of time. The inspectors were concerned about a possible allowance in the TS to have grace from annulus pressure requirements for longer than the allowed LCO required action completion time. Furthermore, a basis for the note and what can be considered ventilating operations was not immediately apparent. Because more information is necessary to evaluate the proper applicability of TS LCO 3.6.15.B and the associated note, future inspection is required to determine if a more than minor performance deficiency or violation exists associated with this issue. Specifically, the inspectors need to determine if a TS compliance issue exists. This is identified as URI 0500390/2015004-04, Shield Building Operability Requirements.
05000346/FIN-2014008-01Davis Besse2015Q2Departure from Method of Evaluation Required Prior NRC Approval Under 10 CFR 50.59 (c)(2)The inspectors identified a Severity Level IV NCV of Title 10, Code of Federal Regulations (CFR) Part 50.59(c)(2), and an associated finding of very-low safety significance for the licensees failure to request and obtain a license amendment pursuant to 10 CFR 50.90. Specifically, the licensees method of evaluation that accepted shield building laminar cracking represented a departure from the method of evaluation described in the Final Safety Analysis Report (as updated), and required prior NRC approval with respect to the design and licensing basis. The licensee entered this finding into its Corrective Action Program; the licensees immediate corrective action determined that shield building remained operable and capable to perform its design safety functions; the licensees planned corrective actions included revising 10 CFR 50.59 Evaluation 13-00918, and preparation of additional documents for inclusion in a license amendment request. The finding was determined to be more than minor because the finding was associated with the Barrier Integrity cornerstone attribute of Design Control, and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609, Appendix A, The SDP for Findings At-Power. Using Exhibit 3, the inspectors determined that the finding screened as very-low safety significance because all the Reactor Containment screening questions for the Barrier Integrity Cornerstone were answered No. Specifically, the inspectors concluded that the shield building remained capable of performing its design safety functions despite the identified laminar cracking. The associated violation was categorized as Severity Level IV because the issue was determined to be of very-low safety significance under the SDP. This finding had a cross-cutting aspect in the area of Human Performance, Conservative Bias, because the licensee did not take a conservative approach to decision making for evaluation of shield building laminar cracking, particularly when information is incomplete or conditions are unusual.
05000346/FIN-2013010-02Davis Besse2014Q3Use of Unqualified Procedure for Ultrasonic Examination of Shield Building RebarThe inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, when the licensee failed to use a qualified procedure for ultrasonic (UT) examination of the Shield Building reinforcing bars (rebar). Specifically, the licensee used a site approved UT examination procedure that had not been qualified to examine the total length of approximately twenty four inches of rebar as specified in the procedure due to near field scanning limitation. The inspectors determined that the performance deficiency of using an unqualified procedure was more than minor and; therefore, a finding 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 have continued use of the unqualified UT examination procedure to examine potential degradation in potentially damaged rebar in the safety-related shield building. Therefore, the licensee could potentially have returned the shield building back to service with unacceptable flaws existing in the rebar. 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. The inspectors answered 'Yes to the questions in Section A of Table 3; and; therefore, the finding was evaluated using the SDP in accordance with IMC 0609, The Significance Determination Process for Shutdown Operations, Appendix G, Attachment 1, Exhibit 4, Barrier Integrity Screening Questions. The inspectors answered all the questions in Exhibit 4 and determined that this finding did not result in degraded physical integrity of the containment during shutdown operations nor did it affect any shutdown safety functions. Therefore, the finding was determined to have very low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, and Evaluation for the licensees failure to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee failed to initially consider the entire length of rebar for potential evaluation and hence, did not consider the appropriate extent of condition.
05000346/FIN-2013010-03Davis Besse2014Q3Failure to Completely Repair Shield Building Concrete VoidingA finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors for the licensees failure in 2011 to properly repair concrete voiding in the shield building that had been identified during that construction opening restoration. The inspectors determined the performance deficiency of failure to completely repair the void during the 2011 shield building restoration was more than minor and; therefore, a finding because the performance deficiency was associated with the Barrier Integrity cornerstone attribute of Design Control and adversely impacted the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the licensees failure to completely repair the concrete voiding in 2011 resulted in the operation of the plant with the shield building in a condition non-conforming to its design basis. The inspectors reviewed the finding using Attachment 0609.04, Initial Characterization of Findings, Table 3 SDP Appendix Router. The inspectors answered No to all the questions in Section A of Table 3 and; therefore, the finding was evaluated using the SDP in accordance with IMC 0609, The Significance Determination Process (SDP) for At-Power Operations, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The inspectors answered all the questions in Exhibit 3 and determined that this finding did not represent an actual open pathway in the physical integrity of reactor containment. Therefore, the finding was determined to have very low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Conservative Bias, for the licensees failure to use decision making practices that emphasize prudent choices over those that are simply allowable. Specifically, the licensee failed to implement a conservative decision to inspect the shield building inside surface void area after repairs had been made during the opening restoration in 2011. Therefore, the licensee missed the opportunity to identify that they had not adequately repaired the void.
05000346/FIN-2013009-01Davis Besse2014Q1Methodology and Acceptance Criteria Utilized for Design and Licensing Basis of the Shield Building with Laminar CrackingThe inspectors identified an unresolved item (URI) regarding the licensees actions to re-establish the design and licensing basis for the SB with identified laminar cracking. Specifically, the inspectors questioned whether the licensee's 10 CFR 50.59 evaluation provided appropriate rationale to support its licensing basis conclusion. The licensee used a combination of testing and calculations to re-establish the design and licensing basis of the SB with laminar cracking. The licensee used additional Impulse Response testing and confirmatory core boring data to more precisely establish the extent of SB laminar cracking. The licensee also performed testing at selected university laboratories to determine rebar splice design capacity in laminarcrack areas. Using these data as input, the licensee performed an evaluation, calculation C-CSS-099.20-063, Revision 0, Shield Building Design Calculation, to demonstrate the SB with laminar cracking had structural capacity to perform its design basis functions consistent with acceptance criteria specified in the design basis code, ACI 318-63, and standard ACI 307-69, referenced in the USAR. Calculation C-CSS 099.20-063 utilized computer software ANSYS to model the shield building and calculate concrete and rebar stress for design basis loading conditions. The inspectors reviewed licensee 10 CFR 50.59 Evaluation 13-00918, related to calculation C-CSS-099.20-063, using NEI 96-07 as guidance. As the licensee described in this evaluation, calculation C-CSS-099.20-063 provided the new evaluation of the shield building, including the effects of laminar cracking for the shield wall, dome, and spring line areas of the building. The calculation included the results of laboratory testing performed to determine the effect of laminar cracking on the structural behavior and strength of the structure. The calculation included a change in methodology. Licensee Evaluation 13-00918 concluded that the licensees use of the ANSYS computer program does not involve a departure from the method of evaluation described in the USAR, because the planned use of ANSYS was considered approved by the NRC for the intended application. Specifically, the licensee compared their use of ANSYS for analytical evaluation of the SB with a similar application of ANSYS reviewed by the NRC and documented in an NRC memorandum dated December 15, 2011, Subject: U.S EPR Design Certification Application Safety Evaluation with Open Items for Portions of Chapter 3, Design of Structures, Components, Equipment and Systems (ADAMS Accession Nos. ML092860252 and ML113081431). The licensee further concluded that a license amendment was not required prior to implementation of the change. The inspectors noted that 10 CFR 50.59 allows a licensee to make changes in the facility as described in its USAR without obtaining a license amendment pursuant to 10 CFR 50.90 only if: (1) a change to the technical specification incorporated in the license is not required, and (2) the change does not meet any of eight criteria specified in that regulation. One of these criteria is specified in 10 CFR 50.59(c)(2)(viii) as result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design basis or in the safety analysis. The inspectors also noted that NEI 96-07, Section 4.3.8, in providing detailed guidance on evaluating changes against that specific criterion, states In general, licensees can make changes to elements of a methodology without first obtaining a license amendment if the results are essentially the same as, or more conservative than, previous results. Similarly, licensees can also use different methods without first obtaining a license amendment if those methods have been approved by the NRC for the intended application. Further, Section 4.3.8.2 in discussing changing from one method of evaluation to another, states that A new method is approved by the NRC for intended application if it is approved for the type of analysis being conducted, and applicable terms, conditions, and limitations for its use are satisfied. Further, licensees are specifically allowed to apply methods that have been reviewed and approved by the NRC, or that have been otherwise accepted as part of another plants licensing basis, without prior NRC approval. That section also provides detailed guidance for determining whether a particular application of a different method is technically appropriate for the intended application, within the bounds of what has been found acceptable to the NRC, and does not require prior NRC-approval. In reviewing Evaluation 13-00918, the inspectors compared the previously approved application referenced by the licensee against the licensees analysis for the shield building laminar cracking. The inspectors agreed that the ANSYS software was capable of accurately modeling the laminar cracking as opposed to the original licensing basis methodology. However, given that the licensees referenced NRC-approved application did not involve modeling of laminar cracking in the structure, the inspectors questioned whether the licensees entire methodology was within the bounds of what has been found acceptable to the NRC. The NEI 96-07 specifies that it is incumbent upon the users of a new methodology to ensure they have a thorough understanding of the methodology in terms of its existing application and conditions/limitations on its use and should document in the 10 CFR 50.59 evaluation the basis for determining it is approved for use in the intended application. In particular, the inspectors questioned whether the application to the shield building laminar cracking, given its uniqueness in the nuclear industry, was sufficiently similar to the referenced NRC-approved application to consider the licensees methodology as NRC-approved or otherwise applied appropriately with respect to the following: It was unclear to the inspectors whether the licensee used an appropriate reference on which to base its conclusion, under provisions of NEI 96-07, that ANSYS was considered "approved by the NRC for the intended application." Specifically, the referenced SER was issued pursuant to an interim phase of the U.S. EPR design certification review process, and hence was not considered a final SER. The inspectors could not identify a licensing action in which the NRC had approved the use of Impulse Response testing and confirmatory core borings to validate the design condition (extent of laminar cracking) that was modeled in the analysis. The inspectors could not identify a licensing action in which the NRC had approved similar licensee laboratory testing to establish/validate rebar splice capacity within laminar crack areas assumed in the analysis. In addition, the licensees test report did not demonstrate the rebar capacity acceptance criteria based on licensee tests were essentially the same as, or more conservative than rebar splice criteria specified in ACI 318-63. The inspectors noted that provisions to check the adequacy of design by use of calculation methods or suitable testing program was a key component of 10 CFR Part 50, Appendix B, Criterion III, Design Control. As noted in NRC IR 05000346/2013004 (ADAMS Accession No. ML13308A283), pursuant to testing conducted for the licensees shield building monitoring program, the licensee in August/September 2013 identified new crack indications some of which may be evidence of crack growth. As a result, the licensee was performing additional testing and analysis with the resulting licensee evaluation currently expected to be completed in mid-2014. Depending on those results, specifically whether cracks were in fact growing and the reason for that growth, it was not clear to the inspectors whether specific monitoring requirements or acceptance criteria with respect to extent of cracking needed to be approved by the NRC for the current operating license. As documented in NUREG-0136, Safety Evaluation Report related to operation of Davis-Besse Nuclear Power Station Unit 1, dated December 1976, the NRC reviewed and accepted ACI 318-63 code provisions that were used in theDavis-Besse safety analysis. Excerpts from NUREG-0136 related to ACI 318-63 include: i. The major code used in the design of concrete seismic Category I structures is ACI 318-63. ii. The design and analysis procedures that were used for these seismic Category I structures are the same as those approved on previously licensed applications and are in accordance with procedures delineated in ACI 318-63 ... and are acceptable. iii. The various seismic Category I structures are designed and proportioned to remain within the limits established by the staff for the various load combinations. These limits are acceptable based on the ACI 318-63 Code ... modified as appropriate for load combinations that are considered extreme. iv. The criteria used in the analysis, design, and construction of all seismic Category I structures to account for anticipated loadings and postulated conditions that may be imposed upon each structure during its service lifetime are in conformance with established criteria, codes, standards, and specifications acceptable to the regulatory staff. The licensee believed its analysis demonstrated the ACI design standard remained valid and that the SB design remained consistent with the standard despite the laminar cracking. However, since the ACI standard did not anticipate or contain provisions to govern evaluation of laminar cracking, the inspectors questioned whether the standard remained applicable/valid for the current condition. Further, it was not clear to the inspectors whether provisions in NEI 96-07 for adopting a different NRC-approved methodology could be used to supplant an industry standard specifically referenced in an NRC SER as the basis for NRC acceptance of the SB design. The inspectors noted a provision in both ACI 318-63 and ACI 349-06, Code Requirements for Nuclear Safety-Related Concrete Structures, endorsed in current regulatory guidance documents, that would appear to allow the licensee to submit its tests and calculation supporting structural adequacy of the shield building with laminar cracking to the NRC for review and approval. This is a process that appears to be distinct from the license amendment process pursuant to 10 CFR 50.90. However, as of the end of this inspection, the licensee had not availed itself of either process to request NRC review and approval of the licensees decision to permanently accept the laminar cracking condition in the shield building wall with respect to the design and licensing basis. This issue is considered an unresolved item pending further review and evaluation by the NRC staff to establish a position on whether the licensee's 10 CFR 50.59 evaluation provided appropriate rationale to support its licensing basis conclusion (URI 05000346/2013009-01, Methodology and Acceptance Criteria Utilized for Design and Licensing Basis of Shield Building with Laminar Cracking). Despite the above design and licensing basis unresolved item, as noted in NRC IR 05000346/2013004, the inspectors continued to believe that the shield building laminar cracking condition remained bounded by the licensees 2011 operability evaluation, and there continued to be reasonable assurance that the shield building remained capable of performing its safety functions.
05000282/FIN-2013002-12Prairie Island2013Q1Licensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, activities affecting quality shall be prescribed by documented instructions, procedures or drawings of a type appropriate to the circumstance and shall be accomplished in accordance with these instructions, procedures or drawings. Contrary to the above, on February 5, 2013, maintenance personnel calibrated a differential pressure switch on the 11 shield building ventilation system, an activity affecting quality, without having instructions, procedures, or drawings appropriate to the circumstance. The failure to have instructions or procedures appropriate to the circumstance resulted in maintenance personnel unknowingly rendering the 11 shield building ventilation system inoperable. In addition, the inappropriate work instructions and a lack of communications from the maintenance personnel to the licensed operators resulted in a failure to implement TS 3.6.9 once the 11 shield building ventilation system became inoperable. The inspectors assessed the significance of this finding using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power. The inspectors determined that this issue was of very low safety significance (Green) because each of the questions contained in the Mitigating Systems portion of IMC 0609, Appendix A, Exhibit 2 could be answered no. The licensee documented this issue as CAP 1369077. Corrective actions for this issue included performing a site wide stand down to reinforce the requirement that work on safety related equipment must be documented by instructions or procedures appropriate for the task at hand, removing the maintenance workers qualifications until remedial training could be provided, and placing an additional supervisor within the specific maintenance department to ensure that individuals fully understood the scope of the work requested to be performed.
05000282/FIN-2012504-01Prairie Island2012Q4Degraded Emergency Action Level SchemeA finding having a significance of preliminarily White with one AV of 10 CFR 50.54(q)(2) associated with risk-significant planning standard 10 CFR 50.47(b)(4) was identified by the NRC for the licensee\\\'s failure to follow and maintain the effectiveness of its emergency plan. Specifically, from July 24, 2011, until May 18, 2012, Prairie Island Nuclear Generating Plants Unit 1 response to the loss of 1R-50 Shield Building Hi Range Vent Gas Radiation Detector failed to restore the capability to classify Emergency Action Levels (EALs), RG1.1, General Emergency, and RS1.1, Site Area Emergency. On May 17, 2012, Corrective Action Program entry 01338120 was written and identified the incorrect repair priority on 1R-50. The instrument was repaired and returned to service on May 18, 2012. This finding was determined to be more than minor because it was associated with the Emergency Response Organization performance attribute of the Reactor Safety Emergency Preparedness Cornerstone. This finding adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. This finding was evaluated in accordance with IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process. As Appendix B was revised in February 2012, the finding was evaluated using both the version in effect at the time of the violation and the current version. Under both versions, other than changing the names of the involved Section and Sheet/Attachment, there was no effect on the final outcome. The issue was determined to be a Failure to Comply. The risk was evaluated using Section 4.0 of IMC 0609 and Sheet 1, Failure to Comply, in the previous revision, and Section 5.0 and Attachment 2, Failure to Comply Significance Logic, in the current revision, along with their associated narratives. With EALs, RG1.1 and RS1.1, ineffective, the inspectors considered mitigating factors, such as alternative EALs, within the same initiating condition and determined the alternative EALs were such that an accurate declaration of the initiating condition would have been made. Therefore, the inspectors determined that no loss of Risk-Significant Planning Standard (RSPS) function existed. However, the alternative EAL classifications would have been delayed, and, therefore, the event would have been declared in a degraded manner. The finding was preliminarily determined to be of low to moderate safety significance (White) in that ineffective EALs, RG1.1, and RS1.1 existed, degraded an RSPS function, and affected the ability of the licensee to properly classify events involving a radiological release. A cross-cutting aspect (H.1(a)) was identified within the decision making component. The licensees risk-significant decision concerning the timely corrective actions to restore the failed 1R-50 Shield Building Hi Range Vent Gas Radiation Detector did not use a systematic process to ensure safety was maintained. A lack of formally defined authority and roles for decisions and communications precluded the appropriate interdisciplinary input, evaluation, and repair of this equipment.
05000305/FIN-2012002-06Kewaunee2012Q1Failure to Submit LER Per 10 CFR 50.73The inspectors identified an SL IV NCV of 10 CFR 50.73(a)(2)(vii) for the failure of the licensee to report an event where a single cause or condition caused two independent trains to become inoperable in a single system designed to control the release of radioactive material. Specifically, the licensee failed to report that both trains of shield building ventilation (SBV) were inoperable due to a single cause, because both trains contained unqualified control card standoffs that were needed to maintain the seismic qualification and operability of the system. The licensee entered this into their CAP as CR429469, planned to perform an ACE, and was drafting an update to Licensee Event Report (LER) 05000305/2011-005. The inspectors determined that the failure to report the event in accordance with 10 CFR 50.73 was a performance deficiency. Because violations of 10 CFR 50.73 are considered to be violations that potentially impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the Reactor Oversight Process (ROP) SDP. Because the performance deficiency, a failure to report, was not an ROP finding per IMC 0612, Appendix B,` Issue Screening,` a cross-cutting aspect was not assigned to this violation. Per the NRC Enforcement Policy, Section The inspectors identified an SL IV NCV of 10 CFR 50.73(a)(2)(vii) for the failure of the licensee to report an event where a single cause or condition caused two independent trains to become inoperable in a single system designed to control the release of radioactive material. Specifically, the licensee failed to report that both trains of shield building ventilation (SBV) were inoperable due to a single cause, because both trains contained unqualified control card standoffs that were needed to maintain the seismic qualification and operability of the system. The licensee entered this into their CAP as CR429469, planned to perform an ACE, and was drafting an update to Licensee Event Report (LER) 05000305/2011-005. The inspectors determined that the failure to report the event in accordance with 10 CFR 50.73 was a performance deficiency. Because violations of 10 CFR 50.73 are considered to be violations that potentially impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the Reactor Oversight Process (ROP) SDP. Because the performance deficiency, a failure to report, was not an ROP finding per IMC 0612, Appendix B,` Issue Screening,` a cross-cutting aspect was not assigned to this violation. Per the NRC Enforcement Policy, Section 6.0,` Violation Examples,` a failure to submit a required LER is categorized as an SL IV violation.
05000346/FIN-2011005-01Davis Besse2011Q4Inadequate Control of Weld Filler Metal ElectrodesA finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings were identified by the inspectors for the licensees failure to control weld rod oven temperature in accordance with procedure WFMC-1 during a rebar splice weld completed for restoration of the shield building access opening. As a corrective action, the licensee removed the welders certification to weld rebar and documented this issue in CR 2011-05536. To ensure that the horizontal rebar splice weld 2H-03R was not affected by delayed hydrogen cracking, the licensees vendor examined the weld splice 48 hours after fabrication and did not identify cracks. The finding was determined to be more than minor because the finding was associated with the Barrier Integrity Cornerstone attribute of Configuration Control and adversely affected the cornerstone objective to provide reasonable assurance that the physical design barriers (e.g., containment) protect the public from radionuclide releases caused by accidents or events. The shield building is part of the containment system. Absent NRC identification, rebar welds would have been fabricated with electrodes exposed to ambient temperatures for excessive periods of time creating a condition that results in hydrogen-induced weld cracking. Rebar splice material with cracks returned to service would increase risk for shield building failure during design basis events such as wind-driven missile impact or earthquake-induced loads. The inspectors completed a significance determination, in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a for the Containment Barrier. Because the issue was corrected promptly, prior to introduction of weld material with hydrogen-induced cracks, the inspectors answered no to each of the four Phase 1 screening questions. Therefore, the finding screened as having very low safety significance. This finding had a cross-cutting aspect in the area of Human Performance, Work Practices because the licensee did not provide adequate supervisory and management oversight of work activities including contractors such that nuclear safety was supported. Specifically, the failure to control the weld rod oven temperature in accordance with procedure WFMC-1 was caused by inadequate licensee oversight of the contracted welder.
05000282/FIN-2011003-12Prairie Island2011Q2Licensee-Identified ViolationTechnical Specification 5.4.1 states that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Section 9 of Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, requires that maintenance that affects the performance of safety-related equipment be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to the above, on February 19, 2011, the licensee failed to properly pre-plan and perform maintenance on the Unit 2 shield building doors with written procedures, documented instructions, or drawings appropriate to the circumstances. Specifically, WO 408666 failed to include information that the performance of maintenance on the Unit 2 shield building doors would render the Unit 2 shield building inoperable. The licensee reported this event to the NRC on April 20, 2011, as required by 10 CFR Part 50.73(a)(2)(v). The licensee also initiated CAP 1271750 to document this issue. Immediate corrective actions included closing the doors to restore shield building operability, revising procedures to clearly state that the opening of both doors can only be done in Mode 5 or 6, and the development of a plant impact statement checklist to aid in determining the impact of maintenance activities on plant operation. The inspectors determined that this finding impacted the Barrier Integrity Cornerstone. The inspectors performed a Phase 1 SDP screening of this issue and determined that it was of very low safety significance (Green) because it did not represent a degradation of the radiological barrier function provided for the control room, the auxiliary building, or the spent fuel pool; did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere; did not represent an actual open pathway in the physical integrity of the reactor containment; and did not involve an actual reduction in the function of the hydrogen ignitors.
05000305/FIN-2011003-04Kewaunee2011Q2Failure to Submit LER per 10 CFR 50.73A Severity Level IV non-cited violation of 10 CFR 50.73(a)(2)(i)(B) and 50.73(a)(2)(v)(C) was identified by the inspectors for the failure of the licensee to report an event or condition that was prohibited by Technical Specifications, and an event or condition that could have prevented the fulfillment of a safety function that is relied upon to control the release of radioactive material. Specifically, the licensee failed to report that shield building ventilation train A was inoperable from December 3, 2010, through January 26, 2011. Technical Specification 3.6.c.1 allows a single train outage time of seven days. Additionally, shield building ventilation train B was inoperable on multiple occasions during the same time period, requiring the licensee to also report an event or condition that could have prevented the fulfillment of a safety function, which is relied upon to control the release of radioactive material. At the end of the inspection period, the licensee was completing an apparent cause evaluation to determine the cause and develop corrective actions. Because violations of 10 CFR 50.73 are considered to be violations that potentially impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the Reactor Oversight Process Significance Determination Process. A cross-cutting aspect was not assigned to this violation. Per the NRC Enforcement Policy, Section 6.0, Violation Examples, a failure to submit a required licensee event report is categorized as a Severity Level IV violation.
05000305/FIN-2011003-03Kewaunee2011Q2Failed Standoffs Result in an Inoperable Train of Shield Building VentilationA 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 failure to have and follow adequate procedures for the evaluation and installation of components in shield building ventilation (SBV) train A. Specifically, the licensee failed to have adequate procedures to direct the completion of a subcomponent classification evaluation (SCE) and prevent nonsafety-related parts from being installed in safety-related applications; have torque specifications for the standoffs (spacers for circuit cards) in the work instructions; and properly accomplish the SCE procedure when evaluating the standoffs. The licensees initial short-term corrective actions removed the installed standoffs from both trains. The licensee also performed an extent-of-condition looking at previously completed item equivalency evaluations to determine if an SCE was needed or missing for newly installed components. The finding was determined to be more than minor because the finding was associated with the Barrier Integrity Cornerstone attribute of procedure quality, and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee failed to have and follow adequate procedures which led to the failure of SBV train A. The inspectors determined that this was a type B containment finding since it was related to a degraded condition that had potential important implications for the integrity of the containment, without affecting the likelihood of core damage. The inspector evaluated the finding using the SDP in accordance with IMC 0609, Appendix H, Containment Integrity SDP, Table 4.1, and determined that the finding did not relate to a containment structure, system, and component, nor containment status that had an impact on large early release frequency. Because of this, the issue screened as Green, using the flowchart in Figure 4.1. The finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to thoroughly evaluate problems such that the resolutions address causes and extent-of-conditions, as necessary. This includes properly classifying, prioritizing, and evaluating for operability and reportability conditions adverse to quality. This also includes, for significant problems, conducting effectiveness reviews of corrective actions to ensure that the problems are resolved. Specifically, the licensee failed to properly evaluate and identify the cause of the SBV train A failure and produce a resolution that addressed the cause (P.1(c)).
05000335/FIN-2010004-05Saint Lucie2010Q3Licensee-Identified ViolationTS 3.0.3 requires that when a limiting condition of operation (LCO) is not met, except as provided in the associated action requirements, within 1 hour action shall be initiated to place the unit in a mode which the specification does not apply. Contrary to this, on July 9, 2007, both trains of shield building ventilation were not operable, and actions were not taken to place the unit in the required mode of operation. This was identified in the licensees CAP as condition report 2010-9892 and Unit 1 LER 02010-004-00. The finding is of very low safety significance because it does not represent an open pathway in the physical integrity of the reactor containment.
05000335/FIN-2010004-04Saint Lucie2010Q3Licensee-Identified Violation10 CFR 50.73, Licensee Event Report System, requires, in part that holders of operating licenses shall submit a Licensee Event Report (LER) within 60 days after discovery of any event such as those described in sections (a)(2)(i)(B), operation prohibited by technical specifications (TS) and (a)(2)(v)(D), a condition that could have prevented the fulfillment a system safety function to mitigate the consequences of an accident. Contrary to this requirement, the licensees discovery of the Unit 1 Shield Building Ventilation System not meeting TS and not being able to fulfill its safety function was not reported within 60 days. The licensee submitted LER 2010- 004-00 on July 16, 2010, exceeding the within 60 day requirement. The inspectors reviewed the issue in accordance with Inspection Manual Chapter 0612 and the NRC Enforcement Manual and determined that traditional enforcement was applicable to the issue because the NRCs regulatory ability was affected. Specifically, the NRC relies on the licensee to identify and report conditions or events meeting the criteria specified in the regulations to perform its regulatory function. The finding was reviewed by NRC management and because the violation was determined to be of very low safety significance, was not repetitive or willful, and was entered into the CAP as condition report 2010-16023, this violation is being treated as a NCV consistent with the NRC Enforcement Policy.
05000390/FIN-2009006-02Watts Bar2009Q2Failure to Follow Plant Procedures for Canceling Preventive MaintenanceA self-revealing NCV of Technical Specification 5.7.1 was identified for the licensees failure to follow plant procedures which resulted in the failure of the Unit 1 Shield Building Vent Radiation Monitor System, an effluent radiation monitor. The inspectors determined the licensees failure to follow site procedures for PM cancellation was a performance deficiency and a finding. The inspectors reviewed Inspection Manual Chapter (IMC) 0612 and determined that the finding is more than minor because the finding is associated with the plant facilities/equipment and instrumentation attribute (reliability of process radiation monitors) of the radiation safety cornerstone (public radiation safety) and adversely affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian use. The finding was assessed using the IMC 0609, Appendix D, Public Radiation SDP, and because there was no failure to implement the effluent program, the finding was determined to be of very low safety significance (Green). No cross-cutting aspect was assigned to this finding because the direct cause was not considered indicative of current performance
05000305/FIN-2008003-02Kewaunee2008Q2Lack of Calculation to Show That the Auxiliary Building Floor Fan Coil Units CAN Perform Their SAFETY-RELATED Function at the Maximum Design Service Water TemperatureThe inspectors identified an unresolved item (URI) due to lack of a calculation to demonstrate that the auxiliary building fan floor FCUs can perform their safety-related function at the maximum design service water temperature of 80XF (degrees Fahrenheit) . Specifically, the inspectors questioned the adequacy of the licensees corrective actions in resolving this issue. This issue is unresolved pending NRC review of the results of the new calculation. The auxiliary building floor contains two FCUs whose safety-related function is to maintain the temperature in the area at 120XF or less. This is the environment qualification temperature for the equipment that is located in this area, categorized as mild environment. These FCUs are cooled by the service water system which has a maximum design inlet temperature of 80XF. Last summer, the licensee commenced a reconstitution of its heating, ventilation and air conditioning calculations. During this effort, the licensee became aware that the calculation used to determine the amount of heat generated in the auxiliary building floor FCU area had several non-conservative assumptions. As an example of these non-conservatisms, the licensee assumed: FnA loss of off-site power (LOOP) during the postulated loss of coolant accident (LOCA) is the most limiting design basis accident for this case. This is a nonconservative assumption as there is additional heat loads generated from nonsafety-related components involved in a non-LOOP LOCA, such as lighting. FnThat the refueling water storage tank (RWST) was full of water; therefore, some heat was exchanged with the tank during this scenario. This might not be the case since during a LOCA, the RWST would empty as the transient develops. FnThat there is some leakage of hot air from the Zone SV (special ventilation) charcoal filters to the environment. The exhaust for this hot air is on the fan floor. The licensee assumed that some of this hot air leaked to the environment but no design basis was found for the number. The licensee performed a test and determined that the actual leakage was much less than what the calculation assumed; therefore, this is a non-conservative assumption. Additionally, the calculation did not take into account the heat generated by the FCU pump motors. All these non-conservatisms questioned the operability of the auxiliary building fan FCUs as well as the other equipment in the area, which is supported by the system including the shield building and special ventilation zone air handling systems. The licensee performed an operability evaluation and included all the non-conservatisms from the previous calculation. The result of this operability evaluation was the FCUs were operable up to a service water inlet temperature of 71XF, but nonconforming with their design requirement of 80XF. Through this inspection cycle, the FCUs and supported systems have remained operable. The licensee is currently performing a more thorough calculation. The inspectors have the following concerns: FnThe service water temperatures may rise to 71XF in approximately the June and July time frame (last year, service water inlet temperature of 77XF was recorded during July). If the licensee fails to prove operability of the FCUs by then, the plant would have to shutdown per TS 3.0.3. FnThe inspectors believe that there is not enough conservatism in the design assumptions to prove operability above 71XF. Currently, the licensee is taking the corrective actions necessary to resolve this issue. This issue is unresolved pending NRC review of the results of the calculation (URI 05000305/2008003-02)
05000305/FIN-2007008-02Kewaunee2007Q2Auxiliary Building Roof DegradationThe inspectors developed an expanded sample scope for the auxiliary building roof degradation that included the review of CAP documents initiated over the past 5 years. The sample selection was made by the inspectors to determine the source of the water that had accumulated in electrical terminal boxes documented in the CAP. Also, the inspectors reviewed, as operating experience, Information Notices (INs) 89-63, Possible Submergence of Electrical Circuits Located Above the Flood Level Because of Water Intrusion and Lack of Drainage, and IN 84-47, Environmental Qualification Tests of Electrical Terminal Blocks. Chronic auxiliary building roof leakage has damaged seals between the auxiliary building and containment. The auxiliary building roof leakage CA was to address the affected component in the leakage pathway, such as the use of caulk, duct tape and plastic barriers and floor berms positioned to collect and direct the water accumulation. The following time line provides documented CAP issues associated with the auxiliary building roof leakage. December 2001 CAP000310 - Aux Bldg Roof Leaking October 2002 CAP013537 - Chronic Auxiliary Building Roof Leak at SW Interface With Containment Building February 2003 CAP014644 - Roof Leaking Near R-21 on Fan Floor in Aux Building June 2003 CAP016785 - Water in EQ Terminal Box 2337 September 2003 CAP017913 - Leak Through Aux Building Roof and Shield Building Wall Joint December 2003 CAP019137 - Auxiliary Building Roof Leak December 2003 CAP019368 - Indications of Chronic Roof Leakage at Sewer Vent Line Above the Aux Bldg 657\' El March 2004 CAP020521 - No CAP Initiated for WR 04-727 (leak dripping onto TB1375) March 2004 CAP020590 - Aux Building Roof Leakage Moving from RCA thru to Clean Area and Back Into RCA April 2004 CAP020846 - Failure to Address Roof Leakage Affecting TB1375 June 2004 CAP021622 - Zone K6 Plant Inspection. Plastic over TB1375 is in poor condition. The duct tape is starting to peel away and the plastic is covered with scale and dirt from water running down it. December 2004 CAP024634 - Auxiliary Building Roof Leak Investigation April 2005 CAP026914 - Found Rusted and Corroded Terminal Strips in EQ Enclosure TB1375 January 2006 CAP031029 - Boot Seal for Penetration 2N Contains Water March 2006 CAP031937 - Water Leaking Into RCA May 2006 CAP033741 - 657\' Elevation. Deteriorated expansion joint between the floor and containment November 2006 CAP039422 - Pull Box Found With Approximately 12 inch of Water in the Bottom. January 2007 CAP040505 - Replacement of the Auxiliary Building Roofing February 2007 CAP041831 - Delaminated / Spalled Shield Bldg. Concrete at El. 664\' of Aux Bldg. Rm. 403 February 2007 CAP041851 - Water Found in Boot Seal at Pen 25N March 2007 CAP043172 - Roof Drain Pipe Leaks at Roof Penetration April 2007 CAP043674 - Auxiliary Building Roof Leak into Electrical Junction Box From the above time line, CAP041831 documented the completion of maintenance rule evaluation MRE003051 on April 27, 2007. The evaluation identified that the condition of the auxiliary building, function 89A-02, should be considered for (a)(1) status. During the inspection, the approval of this recommendation was still under review. The inspectors conducted a walk down of the auxiliary building on May 17, 2007. The inspectors concluded that the chronic nature of the auxiliary building roof leakage resulted in the licensee staff accepting this degrading condition as a housekeeping issue. Specifically, on auxiliary building floor elevation 642, a mop and pail, squeegees and a wet vacuum were positioned for daily clean-up of rain water. The following CAP document excerpts are provide as issue acceptance by the organization. March 2006 CAP031937: The roof of the RCA is leaking. The South west area in the 657\' level of the RCA has had water infiltrating from the outside environment into the RCA for 6 years. The Controlled Area Maintenance Operators (CAMO) and the RP department have had to make removal of the water a routine daily task. April 2007 CAP043674: A roof leak into the 657\' level of the auxiliary building, previously identified by CAP043172, is leaking through an opening in the floor to the 642\' level and wetting an electrical junction box. Recommendation: Redirect leak away from junction box. The inspectors walk down of the auxiliary building also identified two additional issues that the licensee documented in the CAP. The first issue was terminal box number 1376 located on auxiliary building floor elevation 642 that was labeled as environmentally qualified, but did not have a weep hole or a cover to box gasket installed. The licensee generated CAP045012 on this issue. The second issue resulted directly from the auxiliary building roof leakage. Extensive corrosion was identified on the downstream float trap of the B steam generator power operated relief valve (PORV) vent stack drain line. The functionality of the float trap to drain rain water from the PORV vent stack line could not be assured due to possible internal corrosion that may have entered through a corroded trap pipe union. The licensee generated CAP044975 on this issue. The effect of a water volume existing in the downstream side of the PORV required evaluation for impact on PORV lift setpoint and valve functionality during annual environmental temperature changes. At the time of the inspection, this evaluation was not available. As a result, an Unresolved Item (URI) is open pending licensee resolution of the issue. (URI 05000305/2007008-02)
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