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	Q	Significance	CCA	Identified by	Title	Description
05000400/FIN-2007003-02	2007Q2			Licensee-identified	Licensee-Identified Violation	Technical Specification 6.8.1 requires that written procedures be established, implemented and maintained covering the procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978. That appendix discusses procedures for performing maintenance. Contrary to the above, in August, 2005, the licensee failed to implement established maintenance procedures resulting in severe damage to the A reactor water make-up pump impeller and shaft. This failure to implement established maintenance procedures has been entered into the licensees corrective action program (AR 221793). This finding was determined to be of very low safety significance because it did not represent a loss of system safety function.
05000400/FIN-2007402-01	2007Q2			NRC identified	Administration of NRC Required Annual Written Security RE-QUALIFICATION Testing	The inspector identified an Apparent Violation (AV) for two contract security supervisors providing answers to security officers while administering the NRC required annual written security re-qualification test. The finding was contrary to the requirement specified in 10 CFR Part 73 and the Shearon Harris Nuclear Plant, Security Training and Qualification Plan. The inspector determined that the licensees failure to properly supervise the conduct of NRC required annual re-qualification testing is a performance deficiency because the licensee is expected to comply with the Shearon Harris Physical Security Plan. As a result, it cannot be concluded with any degree of certainty that security officers adequately achieved minimum scores which is the criterion used to demonstrate an acceptable understanding of armed security personnel duties. The inspector determined that the finding is greater than minor because it affects the Physical Security Program Objective to prevent radiological sabotage and affects the Cornerstone Attributes for implementation of licensees Protective Strategy and Contingency Response and was reasonably within the licensees ability to correct by ensuring that security officers were being properly tested. The finding was evaluated under the NRCs Enforcement Policy using the traditional enforcement process because the finding involved willfulness and potentially impeded the NRC regulatory process. Additionally, the finding is being considered for escalated enforcement, due to the willful aspects.
05000400/FIN-2007402-02	2007Q2			NRC identified	Administration of NRC Required Annual Plant Access, Radiation Worker and Respiratory Protection Training Testing.	The inspector identified an AV for three contract security supervisors providing answers to security officers while administering annual computer based re-qualification test (Plant Access, Radiation Worker and Respiratory Protection Training). The finding was contrary to the Progress Energy Nuclear Generation Group (Shearon Harris) Standard Procedure TRN-NGGC-0010, Plant Access, Radiation Worker and Respiratory Protection Training, as required by Technical Specification 6.8.1.a and Regulatory Guide 1.33. The inspector determined that the licensees failure to properly supervise the conduct of NRC required annual re-qualification testing is a performance deficiency because the licensee is expected to comply with the Progress Energy Nuclear Generation Group (Shearon Harris) Standard Procedure TRN-NGGC-0010, Plant Access, Radiation Worker and Respiratory Protection Training. As a result, it cannot be concluded with any degree of certainty that security officers adequately achieved minimum scores which is the criterion used to demonstrate an acceptable understanding of plant access and radiation worker procedures. The inspector determined that the finding is greater than minor because it affects the Physical Security Program Objective to prevent radiological sabotage and affects the Cornerstone Attributes for implementation of licensees Protective Strategy and Contingency Response and was reasonably within the licensees ability to correct by ensuring that security officers were being properly tested. The finding was evaluated under the NRCs Enforcement Policy using the traditional enforcement process because the finding involved willfulness and potentially impeded the NRC regulatory process. Additionally, the finding is being considered for escalated enforcement, due to the willful aspects.
05000400/FIN-2008002-01	2008Q1	Green		Licensee-identified	Licensee-Identified Violation	The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV. Technical Specification 3.3.3.6 requires that the B steam generator wide range level transmitter, LT-487, be operable or the plant be shutdown to hot standby with seven days. Contrary to this requirement, LT-487 was not operable since installation of the transmitter connector assembly during initial plant construction because the connector conductors were not properly insulated. Improper insulation of the conductors could have led to an electrical short circuit in the post-accident reactor containment building environment. This finding was determined to be of very low safety significance because alternate indications of adequate heat sink (narrow range steam generator level and auxiliary feedwater flow to the B steam generator) were available during an event which may have caused a harsh environment inside of the reactor containment building. For events that would not cause a harsh environment inside the reactor containment building, LT-487 would function properly as evidenced by its satisfactory operation since it was installed during plant construction. Therefore, the B steam generator was available for core cooling. This event is documented in the licensees corrective action program as AR 221840
05000400/FIN-2008003-01	2008Q2	Green		NRC identified	Failure to Properly Categorize Maintenance Rule Functional Failures	The inspectors identified a non-cited violation (NCV) of 10 CFR 50.65 (a)(2) for the licensees failure to categorize two failures of the condenser vacuum pump effluent radiation monitor (REM-3534) as maintenance rule functional failures and accordingly, failed to monitor the component as required by 10 CFR 50.65 (a)(1). The licensee entered this issue into the Corrective Action Program (CAP) as Condition Report 283579. The finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring the availability, reliability, and capability of systems which responds to initiating events to prevent undesirable consequences. In addition, Example 7.b provided in Manual Chapter 0612, Appendix E, states that violations of Paragraph 10 CFR 50.65 (a)(2), failure to demonstrate effective control of performance or condition and not putting the affected Systems, Structures, and Components (SSCs) in (a)(1), are not minor because they necessarily involve degraded SSC performance or condition. The inspectors determined this finding is of very low safety significance because the REM-3534 is not a risk-significant component and a back-up means of detecting a primary to secondary leak, the steam generator blowdown radiation monitor, was functional during the time periods when REM-3534 was not functional. The finding occurred because of the two missed failures in 2005. All of the failures of REM-3534 since 2005 have been properly counted. Therefore, the cause of this finding was not associated with a cross-cutting area because it is not reflective of current licensee performance. (Section 1R12
05000400/FIN-2008003-02	2008Q2	Green		Licensee-identified	Licensee-Identified Violation	TS 3.6.2.2 requires that the containment spray additive system be operable with two spray additive eductors each capable of adding sodium hydroxide solution from the chemical additive tank to a containment spray system pump flow. Contrary to this, between October 21, 2007 and May 18, 2008 the licensee was unable to maintain proper sodium hydroxide flow in both eductors of the spray additive system. Additional details are located in section 4OA3 of this report. This was identified in the licensees CAP as AR 00254402
05000400/FIN-2008004-01	2008Q3	Green	H.13	Self-revealing	Failure to Maintain Control Over the Station\'s Very High Radiation Areas	A self-revealing Green NCV was identified for the failure to maintain control of access to the stations very high radiation areas (VHRA), as required by 10 CFR 20.1602. The inspectors determined that the licensee failed to maintain sufficient controls of access to VHRAs from the fall of 2006 through January 2008, contrary to 10 CFR 20.1602 and station procedural requirements. Licensee corrective actions included the retrieval and disposition of the security guard master keys, and developing more specific procedural guidance for key control and issuance at the station. The issue was more than minor because it was associated with the Program/Process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure worker health and safety from exposure to radiation by providing the security guards with the means of gaining unauthorized or inadvertent access to areas in which radiation levels could be encountered at 500 rads (5 grays) or more in 1 hour at 1 meter from a radiation source or any surface which the radiation penetrates. The finding was determined to be of very low safety significance because the finding did not involve ALARA planning, collective dose was not a factor, it did not involve an overexposure, there was not a substantial potential for a worker overexposure, and the licensees ability to assess worker dose was not compromised. The cause of the finding was directly related to the risk significant decision making component in the human performance cross-cutting area because of the licensees decision in the Fall of 2006 to create grand master keys which provided security guards unauthorized means of gaining access into VHRAs. (H.1.a) (Section 2OS1
05000400/FIN-2008005-01	2008Q4	Green		NRC identified	Failure to monitor effluent releases from SFP filter backwash	An NRC identified Green non-cited violation of 10 CFR 20.1302 was identified for failure to representatively monitor and assess radioactive effluents from the Spent Fuel Pool Filter Backwash System released via the plant main vent. During the period of approximately July 2000 to April 2007 the plant operated in a configuration that failed to properly implement the effluent monitoring program. With the particulate and iodine monitors being operated without the isokinetic sampling skids, the representativeness of the samples was unknown when the SFP back-wash system was operated. Licensee corrective actions included collection of in-plant samples to bound releases until the monitors are restored to the as designed configuration. The issue was more than minor because it was associated with the Program/Process attribute of the Public Radiation Safety Cornerstone and potentially affected the cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. The effluent release program finding was determined to be of very low safety significance because it was determined that the finding did not involve exceeding regulatory limits in 10 CFR 20.1301(e) or 10 CFR 50 Appendix I, and it was not a substantial failure to implement the effluent program. The finding was determined to not be representative of current operations and the application of a cross-cutting issue was deemed inappropriate
05000400/FIN-2008006-01	2008Q2	Green		NRC identified	Failure to Use Appropriate Acceptance Criteria for Testing Check Valves 1SW-9 and 10	The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, for incorrect test acceptance criteria for Emergency Service Water (ESW) pump discharge check valves 1SW-9 and 10 during test procedure OST- 1214/1215, ESW System Operability Train A/B Quarterly Interval Modes 1-2-3-4-5-6- Defueled. This finding was entered into the licensees corrective action program as condition report NCR 277362. The procedure was immediately placed on hold and planned corrective actions included revision of the ESW pump test procedures to directly observe absence of reverse rotation of the ESW pumps to verify adequate performance of the ESW pump discharge check valves. This finding is more than minor because if left uncorrected, it would become a more significant safety concern since the test procedure could have allowed an inoperable check valve to satisfactorily pass surveillance testing. Specifically, test criteria established would not ensure that the safety objective of preventing pump reverse rotation was achieved. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiency did not result in the ESW pumps being inoperable. (Section 1R21.2.2
05000400/FIN-2008006-02	2008Q2	Green		NRC identified	Failure to Correctly Translate Design Requirements Into Plant Construction Details of Masonry Block Walls	The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to translate critical design attributes into drawings and the as-built condition of the plant. As a result, the fuel and air supplies to the emergency diesel generators (EDGs) were susceptible to risk from impingement due to potential structural failures of a wall for two external events, tornado and seismic. This finding is more than minor because it impacts the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems needed to mitigate the consequences of an accident. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiency, although sufficient to exceed the critical deflection limits and result in cracking, was analyzed to not result in catastrophic collapse. This issue is documented in the corrective action program as NCRs 276674, 277720, and 279326. (Section 1R21.2.9
05000400/FIN-2008008-01	2008Q4	Green		NRC identified	Sprinkler System in Cable Spreading Room A Does Not Meet Licensee\'s Fire Protection Program Requirements	The team identified a non-cited violation of Shearon Harris Unit 1 operating license condition 2.F, for the licensees failure to install the sprinkler system in Cable Spreading Room A (CSRA) in accordance with the approved fire protection program (FPP). Specifically, the installed system would not have been able to deliver the sprinkler system design density of 0.3 gallons per minute/square foot in CSRA, as stated in the FPP in Updated Final Safety Analysis Report Section 9.5.1.2.3. The licensee entered this issue in the corrective action program and established a continuous fire watch in CSRA as a compensatory measure in accordance with the Shearon Harris FPP. The licensees failure to install the sprinkler system in CSRA in accordance with the approved FPP is a performance deficiency. This finding is more than minor because the installed sprinkler system degraded one of the fire protection defense in depth elements and it affected the reactor safety Mitigating Systems cornerstone objective. The team completed a Phase 2 screening of the finding in accordance with IMC 0609, Appendix F, Attachment 1, Part 2, Fire Protection SDP Phase 2 Worksheet, and concluded that the finding was of very low safety significance (Green), in accordance with Step 2.5, Task 2.5.5 of the Worksheet, because there was a safe shutdown path available which was independent of CSRA. The cause of this finding was not associated with a cross-cutting area because it is not reflective of current licensee performance. (Section 1R05.04
05000400/FIN-2008008-02	2008Q4	Severity level Enforcement Discretion		NRC identified	Post-Fire Safe Shutdown From Outside the Main Control Room (Alternative Shutdown)	The team identified a noncompliance of very low safety significance with Shearon Harris Technical Specification 6.8.1.a, for inadequate procedural guidance which directed usage of instruments that were not protected from fire damage in FZ 12- A-6-PICR1. Specifically, procedure AOP-004, Remote Shutdown, directed the operators to verify emergency service water (ESW) header flows were above the minimum flow requirement of 7500 gpm using flow indicators (FI) FI-9101A2 and FI-9101B2. These instruments would be unreliable during operation from the ACP because their cables were routed through the FZ of concern and the cables were not protected from fire damage. The violation meets the criteria of NRC Enforcement Policy, Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) for enforcement discretion. The team noted that procedure AOP-004, Remote Shutdown, would be used to safely shut down the plant from the ACP (utilizing Train B equipment) for a fire in FZ 12-A-6-PICR1. The procedure directed operators to verify ESW header (HDR) flows using FI-9101A2, A HDR Flow and FI-9101B2, B HDR Flow. The team reviewed cable routing data and noted that FI-9101A2 and FI-9101B2 may not provide reliable ESW flow indication for the operators at the ACP because the cables were routed through FZ 12-A-6-PICR1 and were not protected from fire damaged. This may potentially delay operator actions required to bring the plant to SSD conditions. Based on discussions with operations personnel and review of service water system simplified flow diagrams, the team determined that the ESW system was flow balanced to ensure that the 7500 gpm minimum flow would be provided to ESW HDR A and ESW HDR B. During walkdowns of the ACP, the team noted that valve position indication (i.e., open/close) was provided at the ACP for various ESW valves, including 1SW-270, HDR A to Auxiliary Reservoir and 1SW-271, HDR B to Auxiliary Reservoir. These valves were required to be opened (from the ACP) to ensure adequate ESW header flow. Review of cable routing data for valve 1SW-271 showed that this valve was not routed through FZ 12-A-6-PICR1 and would not be affected by a postulated fire in this FZ. The team determined that, based on operator experience, training, and indication of the position of ESW valve 1SW-271 at the ACP, it was likely plant operators would be able to determine that sufficient ESW flow was available and they would take the appropriate actions required to ensure post-fire SSD conditions. The licensee initiated NCR 298072 to address this issue in the CAP.
05000400/FIN-2009003-01	2009Q2	Green		NRC identified	Review the Cooling Tower Blowdown Line Pathway Dose Compared to Doses from all other pathways	The inspectors identified an unresolved item associated with the leakage of radioactive liquid effluents into the ground from cracks in the CTBL. This item is unresolved pending further review and evaluation of the licensees final dose assessment for the CTBL pathway. The inspectors reviewed with licensee representatives the licensees vendor report regarding the assessment and evaluation of the increase in tritium identified in ground water samples wells along the CTBL as documented in AR #00309035. The licensee discharges permitted and monitored radioactive liquid effluents into the CTBL for dilution with a release point into the Harris Lake. On December 15, 2008, the licensee had observed water in Air Relief System Manhole (ARSM) Number (No.) 2 located on the CTBL upstream from the permitted release point. The licensee obtained water samples from ARSM No. 2 for analysis and identified tritium levels ranging from less than the detection limit to 2,120 picoCuries per liter (pCi/L). As a result, the licensee conducted a hydrology report and assessment of the CTBL. From that assessment the licensee installed nine groundwater monitoring wells at various points along the CTBL and ARSM No. 2 from January 21 March 4, 2009. At the time of the onsite inspection, the licensee had collected several monthly samples with tritium levels ranging from less than the detection limit to 2,450 pCi/L. Some wells were found to be dry. At the time of the onsite inspection the licensee was still evaluating the results of the groundwater monitoring wells. The licensee also evaluated the inside of the CTBL. The licensee identified numerous cracks and plant roots growing into the CTBL. In addition, there was approximately 3,000 feet of the CTBL (located downstream of the Cooling Tower but upstream from ARSM No. 2) that was not evaluated due to worker safety conditions (e.g., slippery conditions due to mud, low oxygen concentrations, etc.). At the time of the onsite inspection, the vendor had not submitted its final assessment and evaluation report of the CTBL to the licensee. As a result, the licensee had not evaluated and assessed the amount of radioactive liquid effluents released into the ground from cracks in the CTBL. An unresolved item (URI) was identified regarding the significance of the CTBL leakage pathway with regard to meeting the requirements of the Offsite Dose Calculation Manual (ODCM). The ODCM states that radioactive materials released in liquid effluents to unrestricted areas are required to demonstrate compliance with 10 CFR 50 Appendix I. The calculated annual total quantity of all radioactive materials above background to be released from each light-water-cooled nuclear power reactor to unrestricted areas will not result in an estimated annual dose or dose commitment from liquid effluents for any individual in an unrestricted area from all pathways of exposure in excess of 3 millirems to the total body or 10 millirems to any organ. The dose commitment had not been determined due to an unevaluated release pathway where releases were occurring at a location other than designed. Specifically, radioactive liquid effluents were being released into the ground from cracks in the CTBL. In accordance with the ODCM, the liquid effluent release point is at the point of discharge from the CTBL into Harris Lake. This item is unresolved pending NRC review and evaluation of the final dose assessment for the CTBL pathway. URI 05000400/2009003-01, Review the Cooling Tower Blowdown Line Pathway Dose Compared to Doses from All Other Pathways
05000400/FIN-2009003-02	2009Q2	Green	H.8	Self-revealing	Failure to Provide Procedures to Control and Adjust the Manipulator Crane Gear Limit Setpoints	A self-revealing Green NCV of Technical Specification (TS) 6.8.1, Procedures, was identified when the licensee failed to follow Attachment 4, Manipulator Crane and Auxiliary Hoist Checkout, of Fuel Handling Procedure 20 (FHP-020), Refueling Operations, resulting in damaged grid straps on two fuel assemblies on April 23, 2009. Specifically, the value of the manipulator crane gear limit setpoints for the lower core slow zone exceeded the values allowed by the checkout procedure. This resulted in the fuel handlers damaging the grid straps on two successive fuel assembly moves. The licensee entered this issue into their corrective action program (CAP) as action request (AR) #332368. As corrective actions, the licensee suspended the core offload, reset the lower core slow zone within tolerance, and permanently discharged the affected fuel assemblies. Additionally, the licensee committed to revise FHP-020 prior to the next refueling outage in order to prevent recurrence. The violation was more than minor because it is associated with the human performance attribute of the Barrier Integrity cornerstone, and it affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The finding was determined to be of very low safety significance because it was a deficiency associated with fuel handling errors that did not cause damage to fuel clad integrity or a dropped fuel assembly. The finding has a crosscutting aspect of Procedural Compliance, as described in the Work Practices component of the Human Performance cross-cutting area because the licensee accepted the out of tolerance values that were outside the acceptance criteria of the procedure (H.4(b)). (Section 4OA2
05000400/FIN-2009003-03	2009Q2	Green		Licensee-identified	Licensee-Identified Violation	Technical Specification 6.12. a requires, in part, that each high radiation area, in which the intensity of radiation is > 100 mrem/hr but < 1000 mrem/hr, measured at 30 cm from the radiation source or from any surface the radiation penetrates, shall be barricaded and conspicuously posted as a high radiation area. Contrary to the above, on March 23, 2009, the licensee constructed scaffolding that provided a potential access to the Filter Backwash Transfer Tank room that was not conspicuously posted or barricaded. Licensee evaluations performed after the event showed that the intensity of radiation was >100 mrem/hr but <1000 mrem/hr measured at 30 cm from the pipe surfaces in those areas. This finding was entered in the licensees corrective action program March 25, 2009 as AR #327372. This finding is of very low safety significance because there was no evidence of unauthorized worker entry into the area and no unexpected /unintended radiation exposures to licensee personnel
05000400/FIN-2009003-04	2009Q2	Green		Licensee-identified	Licensee-Identified Violation	Technical Specification 3.3.2 requires, in part, a minimum of two operable channels of Loss of Offsite Power primary undervoltage (UV) protection on each 6.9 kV Emergency bus. Contrary to this, between January 26, 2006 and January 26, 2009 the licensee failed to maintain the proper 1A-SA Emergency Bus primary UV trip setpoint. This finding was assessed using the Phase 1 screening worksheet of the SDP (Attachment 4 of Manual Chapter 0609, Significance Determination Process) and determined to be of very low safety significance because the 1A-SA Emergency Bus and A EDG would have still performed their safety function even though they would have been delayed by approximately 1 msec. Additional details are located in section 4OA3 of this report. This was identified in the licensees CAP as AR #316381.
05000400/FIN-2009004-01	2009Q3	Green	H.2	NRC identified	Failure to Maintain an Adequate Quality Assurance Training Program	The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the licensees failure to maintain an adequate training program for personnel performing activities affecting quality as necessary to assure that suitable proficiency is achieved and maintained. The licensees training program was inadequate because the means to maintain QC inspector proficiency and the QC continuing training program failed to ensure that QC inspectors employed appropriate inspection techniques. This failure was manifested in three separate quality control electrical verification errors during plant modifications made in April and May 2009. The licensee entered this issue into their CAP as action request (AR) #341355. As corrective action, the licensee correctly reinstalled and verified the modifications to be in accordance with plant design. Additionally, the licensee committed to revise and/or create procedures to institutionalize QC training in an initial training and certification program, as well as a continuing training program. This violation was more than minor because if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. This finding is associated with the Design Control attribute of the Mitigating Systems cornerstone, and it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using Attachment 4 of IMC 0609, the significance of this finding was determined to be of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of system safety function, did not represent actual loss of safety function of a single train for longer than its Technical Specification (TS) Allowed Outage Time, did not represent an actual loss of safety function of one or more non-TS Trains of equipment designated as risk-significant, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a crosscutting aspect of Supervisory and Management Oversight, as described in the Work Practices component of the Human Performance cross-cutting area because the lack of oversight and engagement by management resulted in the inadequate QC training program (H.4(c)). (Section 4OA2
05000400/FIN-2009005-01	2009Q4	Green		NRC identified	Failure to Properly Install Spot Type Smoke Detectors	The inspectors identified a Green NCV of the Shearon Harris Nuclear Power Plant Operating License condition 2.F, Fire Protection Program, for failing to correctly install spot type smoke detectors between four and twelve inches down from the ceiling to the top of the detector as required by National Fire Protection Association (NFPA) 72E, Automatic Fire Detectors. Specifically, it was determined that eight spot type smoke detectors are installed approximately five feet below the ceiling in the plants Computer Room. The licensee took immediate corrective action by initiating compensatory fire watches. The licensee entered this into the corrective action program (CAP) as Action Request (AR) #363555. The finding was determined to be more than minor because it affected the Mitigating Systems Cornerstone objective of availability, reliability, and capability of the fixed fire detection system and was associated with the protection against external factors (fire) attribute. Specifically, this failure could affect the timeliness of response to a fire due to the delayed detection of smoke and resulting alarm, allowing the fire to grow larger prior to the fire brigade taking action. Using MC 0609, Appendix F, it was determined that this issue was in the category of fixed fire protection systems which had moderate degradation due to the fact that the system would function, although delayed. Further, it was determined that this issue was a Fire Damage State Zero (FDS0). As such, only the fire ignition source and initiating fuels are damaged by the fire. FDS0 is not analyzed in the fire protection SDP as a risk contributor and is therefore of very low safety significance (Green). Due to the fact that this condition has been present since initial installation during plant construction, it was determined that this was not indicative of current licensee performance and therefore no cross-cutting aspect was identified
05000400/FIN-2009005-02	2009Q4			NRC identified	A ESW Pump Power Supply Cables Submerged in Water	The inspectors identified an unresolved item (URI) associated with the submergence of Safety Related cables in an underground bunker. This item is unresolved pending further review and evaluation of the licensees environmental qualifications of submerged 6.9kV cabling. The inspectors identified A ESW pump power supply cables submerged in approximately 2.5 feet of standing water within manhole MH73B-SA. Additional inspection activities are needed to determine if the A ESW pump power supply cables are suitable for submersion in water. Pending the results of this additional inspection, an URI will be opened and designated as URI 05000400/2009005-02, A ESW Pump Power Supply Cables Submerged in Water
05000400/FIN-2009006-01	2009Q4	Green	P.2	NRC identified	Failure to Preclude Repetition of a Significant Condition Adverse to Quality for Both Containment Spray Additive System Eductors Being Outside of the Technical Specification Flow Band	The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, \"Corrective Action,\" for the licensees failure to identify the cause and take corrective actions to preclude repetition of a significant condition adverse to quality for both containment spray additive system eductors being outside of the technical specification flow band. Specifically, between July 2009 and the present, the violation occurred when Eductor A was found three times and Eductor B was found once outside of the Technical Specification 3.6.2.2 flow band. This issue was previously identified as a significant condition adverse to quality in January 2008, but the corrective actions taken failed to preclude repetition. The licensee entered this issue into the corrective action program as nuclear condition report 356873. The licensee took immediate corrective actions to throttle the eductor flow to within the band, and is developing corrective actions to preclude repetition. The finding is more than minor because it is associated with the design control attribute of the Barrier Integrity Cornerstone and affects the cornerstone objective of providing reasonable assurance that physical design barriers, such as the iodine scrubbing capability of the containment spray additive system eductors, will protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" the finding was determined to have a very low safety significance because it did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, or spent fuel pool; the finding did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere; the finding did not represent an actual open pathway in the physical integrity of reactor containment; and the finding did not involve an actual reduction in function of the hydrogen igniters in the reactor containment. The finding had a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary, and for significant problems, conduct effectiveness reviews of corrective actions to ensure that the problems are resolved (P.1(c)) (Section 4OA2.a(3)(i))
05000400/FIN-2009006-02	2009Q4	Green	H.14	NRC identified	Failure to Correct a Condition Adverse to Quality Involving a Main Steam Isolation Valve Degrading Trend Before Valve Failure	The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, \"Corrective Action,\" for the licensees failure to correct a condition adverse to quality in a timely manner. Specifically, between May 27, 1997 and September 29, 2007, Main Steam Isolation Valve 82 close stroke time exhibited a condition adverse to quality for a trend degrading towards the technical specification limit, without sufficient corrective actions to prevent failure. This resulted in Main Steam Isolation Valve 82 exceeding the five-second stroke time limit required in Technical Specification 3.7.1.5. The licensee entered this issue into the corrective action program as nuclear condition report 358464. This finding is more than minor because it is associated with the containment barrier performance attribute of the Barrier Integrity Cornerstone and affects the cornerstone objective of providing reasonable assurance that physical design barriers, such as the main steam isolation valve radiological release barrier required for a steam generator tube rupture, protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" the finding was determined to have a very low safety significance because it did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, or spent fuel pool; the finding did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere; the finding did not represent an actual open pathway in the physical integrity of reactor containment; and the finding did not involve an actual reduction in function of the hydrogen igniters in the reactor containment. This finding had a cross-cutting aspect in the area of human performance associated with decisionmaking because the licensee did not use conservative assumptions so that safety-significant decisions were verified to validate underlying assumptions and identify unintended consequences (H.1.(b)) (Section 4OA2.a(3)(ii))
05000400/FIN-2009006-03	2009Q4			NRC identified	Unresolved Item Associated with the Evaluation of the Failure of Emergency Service Water Valve Failure	The inspectors identified a URI associated with the evaluation of the failure of ESW Auxiliary Reservoir Discharge Valve 271 to open on the start of ESW Pump B. Description. On October 19, 2007, while in Mode 5, ESW Auxiliary Reservoir Discharge Valve 271 failed to open on the start of ESW Pump B. This valve is required to open on the start of an ESW pump to provide a discharge path for the cooling water. Operators immediately stopped ESW Pump B and aligned normal service water to the safety related components in Train B. The licensee determined that the auto open controls for Valve SW-271 had been disabled by a clearance order for unrelated work. Although ESW Train B is not required to be operational in Mode 5, the components cooled by ESW Train B, such as EDG B and RHR Train B, were being relied upon as protected train equipment. Therefore, ESW Train B was necessary to ensure core decay heat removal in the event that off-site power was not available. NRC inspectors wrote a self-revealing NCV of TS 6.8.1, Programs and Procedures, for an inadequate clearance order as documented in NRC Integrated Inspection Report 05000400/2007005. The team reviewed the evaluation performed for this NCV including the reportability review. The reportability review stated this condition was not reportable since operators were able to open this valve manually from the control room. The team questioned whether the operators would be able to open the valve within one minute, which is required to ensure cooling to the EDGs during an accident. The team also determined that when the valve is manually opened by the reactor operators from the control room, that the valve would automatically go closed due to the inadequate clearance. As a result of the teams questions, the licensee wrote NCR 358062 and determined that the failure of SW-271 to open was a MRFF. This failure did not exceed the ESW Train B maintenance rule performance criteria. The licensee determined that this failure affected the MSPI. This condition could prevent the fulfillment of the safety function of EDG B and RHR B that are needed to maintain the reactor in a safe shutdown condition or to remove residual heat. The licensee wrote NCR 361821 to address this issue. This issue is considered unresolved pending additional NRC review of the evaluation of the failure including the reportability review, the risk assessment, and the corrective actions: URI 05000400/2009006-03, Unresolved Item Associated with the Evaluation of the Failure of Emergency Service Water Valve 271
05000400/FIN-2009201-01	2009Q2	Green	P.2	NRC identified	Security
05000400/FIN-2009201-02	2009Q2	Green		NRC identified	Security
05000400/FIN-2009201-03	2009Q2	Green		NRC identified	Security
05000400/FIN-2010002-01	2010Q1	Severity level IV	P.2	NRC identified	Failure to Submit a Licensee Event Report for a Condition Prohibited by Technical Specifications Associated with the B Emergency Service Water Discharge Valve	The inspectors identified a Severity Level IV, non-cited violation (NCV) of 10CFR 50.73(a)(2)(i)(B) due to the licensees failure to recognize that the inability of the B Emergency Service Water (ESW) Discharge Valve (1SW-271) to open on the start of B ESW pump caused a reportable condition. Consequently, the licensee failed to submit a licensee event report (LER) within 60 days as required by 10 CFR50.73. The licensee entered this issue into the corrective action program (CAP) as Action Request (AR) #361821 and AR #358062. The licensee took corrective action by reporting this event in LER 05000400/2010-001, Clearance Error Results in Equipment Becoming Inoperable. The licensees failure to recognize that the inability of 1SW-271 to open caused a reportable condition and submit an LER as required by 10 CFR 50.73 was a performance deficiency. This issue was dispositioned as traditional enforcement, instead of the Significance Determination Process, because it had the potential for impacting the NRCs ability to perform its regulatory function. However, because this violation was of very low safety significance, was not repetitive or willful, and was entered into the licensees CAP as AR #361821 and AR #358062, the NRC has characterized the significance of this violation as a Severity Level IV NCV in accordance with section IV.A.3 and supplement I of the NRC Enforcement Policy. The cause of this event was directly related to the cross-cutting aspect in the area of problem identification and resolution within the CAP component because the licensee did not adequately evaluate the need to submit an LER per the requirements of 10 CFR 50.73.
05000400/FIN-2010002-02	2010Q1	Green	P.5	NRC identified	Failure to Promptly Evaluate Operating Experience and Identify Potential Steam Voiding as a Condition Adverse to Quality	The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly evaluate operating experience (OE) received October 22, 2008 and identify potential steam voiding in the residual heat removal (RHR) system as a condition adverse to quality. During the evaluation, which was not completed until July 16, 2009, the licensee learned that the suction lines for the RHR pumps are susceptible to steam voiding at temperatures as low as 240F. If the steam void flowed to an RHR pump, that pump could fail causing the associated train of the Emergency Core Cooling System (ECCS) to fail. The delay in evaluating the OE resulted in a delay of determining and implementing appropriate corrective actions. Specifically, the failure to promptly evaluate this OE enabled the licensee to violate Technical Specification (TS) 3.0.4when the plant transitioned from Mode 4 to Mode 1 with only one operable train of ECCS after refueling outage (RFO) 15 on May 9, 2009. The licensee entered this issue into the CAP as AR #345425. The licensee took corrective action by changing procedures to avoid exposing the suction lines to excessive temperatures during Modes when it is required to be operable for ECCS, thereby preventing potential steam voiding. The inspectors determined that the failure to promptly evaluate OE received on October 22, 2008, and identify potential steam voiding as condition adverse to quality was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it could have potentially caused one or more RHR pumps and associated ECCS trains to be inoperable due to steam voiding. Using Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors concluded that a Phase 2 evaluation was required because this finding represented a potential loss of safety function of the RHR system. The inspectors performed a Phase 2 analysis using IMC 0609Appendix A, Determining the Safety Significance of Reactor Inspection Findings for At-Power Situations and the site specific risk informed inspection notebook. Due to the site specific risk informed inspection notebook not containing appropriate target sets to accurately estimate the risk input of the finding, it was determined that a Phase 3 analysis was required. A regional Senior Reactor Analyst performed the Phase 3 evaluation and concluded the finding was of very low safety significance (Green). The NRCs most current Standardized Plant Analysis Risk Model was used for the evaluation. The evaluation assumed that the B RHR Pump always failed to start for the exposure time of seventy hours. Also, there was a potential increase in the common cause failure of the RHR pumps. The dominant accident sequence was a postulated Small Break LOCA with initial success of the ECCS via High Pressure Injection, but the ECCS failed in the recirculation mode. The SDP performed for this violation considered the potential loss of safety function of the RHR system and therefore bounded all violations described in LER 05000400/2009-002 which is further discussed in Section 4OA3.2.This finding was determined to have a cross-cutting aspect in the OE component of the Problem Identification and Resolution area, in that the licensee failed to evaluate OE in a timely manne
05000400/FIN-2010002-03	2010Q1	Green	P.3	NRC identified	A ESW Pump Power Supply Cables Submerged in Water	The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, in that the licensee failed to maintain the A ESW pump power cables in an environment for which they were designed. Specifically, the cables were submerged in water in manway 73B-SA, a condition for which they were not qualified. The licensee entered this issue into the CAP as AR #376709. As immediate corrective action, the licensee pumped the manway dry. The inspectors determined that the failure to ensure that the A ESW pump power cables were maintained in an environment for which they were designed was a performance deficiency. The finding was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, it could have caused the A ESW pump to become inoperable in the event that the cable failed due to long term degradation as a result of continuous submergence. The finding affected the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the significance of this finding using IMC0609, Significance Determination Process, Phase 1 Worksheet. The finding was of very low safety significance because it was a qualification deficiency that did not result in a loss of operability. This finding was determined to have a cross-cutting aspect in the CAP component of the Problem Identification and Resolution area associated with timely and effective corrective actions
05000400/FIN-2010003-01	2010Q2	Green	H.11 H.12	Self-revealing	Failure to Follow Procedure to Install the Load Block 5 Auxiliary Relay	A self-revealing Green NCV of TS 6.8.1, Procedures, was identified for the licensees failure to follow procedure PIC-E069, Sequencer Electomechanical Timing Relays; D.C. Pick-Up, D.C. Drop-Out, A.C. Pick-Up, and A.C. Drop-Out. Specifically, the licensee failed to properly reinstall the Load Block 5 Auxiliary Relay, resulting in the automatic start of B Motor Driven Auxiliary Feedwater (MDAFW) pump and water flowing to all three steam generators. Operators immediately secured the B MDAFW pump. The licensee entered this issue into their corrective action program (CAP) as action request (AR) #381672. As corrective action, the licensee removed and correctly installed the relay followed by a successful post maintenance test. Additionally, the licensee plans to revise ADM-NGGC-0104, Work Management Process, to require the work implementer to specify which mitigating actions and/or human performance barriers will be used for critical steps. The failure to follow procedure PIC-E069 section 7.6 for the restoration of the load block 5 auxiliary relay was a performance deficiency. The violation was more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone, and it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, it resulted in the automatic start of the B MDAFW pump and water flowing to all three steam generators. Using IMC 0609, Significance Determination Process, Phase 1 screening worksheet of the SDP this finding was determined to be of very low safety significance because it was not a design or qualification deficiency confirmed to result in a loss of operability or functionality, did not represent a loss of system safety function, did not result in a loss of safety system function for a single train for greater than TS allowed outage time, did not result in a loss of safety function of one or more non-TS trains of equipment designated as risk-significant for greater than 24 hours, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect of Human error prevention, as described in the Work Practices component of the Human Performance cross-cutting area because the licensee did not apply sufficient human error prevention tools to ensure the correct installation of the relay (H.4(a))
05000400/FIN-2010003-02	2010Q2	Green	H.5	Self-revealing	Reactor Trip due to Failing to Properly Assemble an Oil Filter in the Hydrogen Seal Oil System	A self-revealing Green finding was identified for the licensees failure to follow Work Control Management procedure WCM-006, Graded Approach to Planning and Scheduling, which has requirements that would have ensured the proper rebuild of the oil filter assembly in the hydrogen seal oil (HSO) system. Specifically, this resulted in inadequate maintenance on the filter assembly which caused the handle of the assembly to eject during power operations, causing an oil spill which necessitated a manual reactor trip. The licensee entered this issue into the CAP as Action Request (AR) #366174. The licensee took corrective action to replace the oil filter assembly, as well as clean and replace the spilled oil. Additionally, the licensee reviewed both completed and upcoming work orders to verify they were properly classified based upon potential impact on plant operations. The licensees failure to follow WCM-006 requirements which resulted in the improper rebuild of the oil filter assembly in the HSO system was identified as a performance deficiency. The finding was determined to be more than minor because it was associated with the procedure quality attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, the performance deficiency resulted in an initiating event causing a manual reactor trip and the possibility of an oil fire in the vicinity of the offsite power electrical supply ducts. Using IMC 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors concluded that a Phase 2 evaluation was required since the finding contributed to both the likelihood of a reactor trip and the likelihood that mitigating systems would not have been available. This conclusion was based upon the potential for the spilled oil to ignite in a location that could have challenged the offsite electrical power supply bus ducts following the reactor trip. A regional Senior Reactor Analyst completed a Phase 3 evaluation under the Significance Determination Process. The performance deficiency was characterized as of very low safety significance (Green) based upon the results of this evaluation. The dominant accident sequence involved the postulation of oil igniting in the spill zone. Once ignited, suppression efforts were unsuccessful, causing the loss of the turbine building and a loss of offsite power. Given this damage state, recovery of offsite power was not considered credible. Subsequently, it was postulated that the emergency diesel generators failed which ultimately led to a loss of core cooling and core damage. The finding has a cross cutting aspect of Work Planning, as described in the Work Control component of the Human Performance cross-cutting area because the failure to correctly classify the work package as Quality Critical resulted in not correctly mitigating the risk associated with working on this equipment by including additional guidance to assist the technicians in completing the work successfully (H.3(a)) (Section 4OA3).
05000400/FIN-2010005-01	2010Q4	Green	H.9 H.2(b)	NRC identified	Failure to Properly Implement Procedural Guidance to Maintain the FHBEES Boundary	The inspectors identified a Green NCV of Technical Specification (TS) 6.8.1, Procedures, for the licensees failure to properly implement procedural guidance to maintain the Fuel Handling Building Emergency Exhaust System (FHBEES) boundary. Specifically, the licensee failed to properly implement procedural guidance to maintain the FHBEES boundary while two doors were propped open on October 21, 2010 and October 22, 2010. This was apparent when the inspectors identified one individual unaware of their responsibilities and another individual inattentive. The licensee entered this issue into their CAP as action request (AR) #428580 and AR #428858. The licensee took corrective action to relieve the inattentive individual and conducted additional training for all of the other individuals responsible for closing the doors. The failure to properly implement procedural guidance to maintain the FHBEES boundary while two doors were propped open from October 21, 2010 until October 22, 2010 was a performance deficiency. The performance deficiency was more than minor because it was associated with the Barrier Performance attribute of the Barrier Integrity cornerstone 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 potential safety consequence is that if spent fuel had been damaged in the spent fuel pool during this time, the FHBEES may not have been able to properly filter and monitor a radioactive release. Using IMC 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined this issue to be of very low safety significance because it only represented a degradation of the radiological barrier function provided for the fuel handling building. The finding has a cross-cutting aspect of Training and Work Hours, as described in the Resources component of the Human Performance crosscutting area because the licensee did not effectively train the individuals regarding their procedural responsibilities when the FHBEES doors were propped open (H.2(b)).
05000400/FIN-2010005-02	2010Q4	Green	H.11 H.12	Self-revealing	Failure to Follow Procedure Results in Emergency Safeguards Sequencer Actuation and Safety Injection Signal (SIS) while the Plant was in Mode 6	A self-revealing Green NCV of Technical Specifications (TS) 6.8.1, Procedures, was identified for the licensees failure to follow procedure MST-I0073, Train B 18 Month Manual Reactor Trip, Solid State Protection System Actuation Logic & Master Relay Test. Specifically, step 7.4.14 of MST-I0073 required the licensee to place the Master Relay Selector Switch (MRSS) in the Off position. Contrary to this requirement on October 28, 2010, the licensee failed to place the MRSS in the Off position at step 7.4.14. Instead, at step 7.5.85, the technicians noticed that the MRSS remained in Position 3 and then placed the MRSS in the Off position. This action combined with the current plant condition caused an invalid B train safety injection signal (SIS) and B Emergency Safeguards Sequencer (ESS) actuation while the plant was in Mode 6. The licensee entered this issue into their corrective action program (CAP) as action request (AR) #430289. As corrective action, the licensee restored the plant to the pre-actuation condition and conducted training for the maintenance technicians. The failure to follow procedure MST-I0073 for the proper operation of the MRSS was a performance deficiency. The finding was more than minor because it is similar to the more than minor example 4.b from MC 0612 Appendix E in that an operator incorrectly operated a switch causing a plant transient. Additionally, it is associated with the human performance attribute of the Mitigating Systems cornerstone, and it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, it resulted in an invalid SIS causing the ESS to start the B ESW and B CCW pumps. Using IMC 0609, Significance Determination Process, Phase 1 screening worksheet and Appendix G (Shutdown Operations), Attachment 1, Checklist 4, this finding was determined to be of very low safety significance because it did not meet any of the guidelines which require quantitative assessment. The finding has a cross-cutting aspect of Human Error Prevention, as described in the Work Practices component of the Human Performance cross-cutting area because the technicians proceeded in the face of uncertainty without consulting supervision when they encountered unexpected plant conditions (H.4(a)).
05000400/FIN-2010005-03	2010Q4	Green	H.14	NRC identified	Failure to comply with the limiting conditions for operation, while the Refueling Water Storage Tank was aligned to the non-seismically qualified Fuel Pool Purification System	The inspectors identified a Green NCV of TS 3.1.2.6, Borated Water Sources, for the failure to comply with the limiting conditions for operation, while the Refueling Water Storage Tank (RWST) was aligned to the non-seismic Fuel Pool Purification system (FPPS) for purification, causing the RWST to be inoperable. Specifically, when FPPS was aligned to the RWST, the licensee did not declare the RWST inoperable. The licensee took corrective actions (AR #422180) and revised OP-116.1, FPPS, to remove the capability to purify the RWST in Modes 1 through 4. The failure to comply with the actions of TS Limiting Condition for Operation (LCO) 3.1.2.6 while the Refueling Water Storage Tank (RWST) was aligned to the nonseismic FPPS for purification on May 24, 2010, causing the RWST to be inoperable, was a performance deficiency. The performance deficiency was more than minor because it affected the Design Control attribute of the Mitigating System cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, when the FPPS was aligned to the RWST, the licensee did not declare the RWST inoperable. The inspectors evaluated the significance of this finding Using Attachment 4 of IMC 0609, the significance of this finding was determined to be of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of system safety function, did not represent actual loss of safety function of a single train for longer than its TS Allowed Outage Time, did not represent an actual loss of safety function of one or more non-TS Trains of equipment designated as risk-significant, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect of Conservative Assumptions, as described in the Decision Making component of the Human Performance cross-cutting area because, assumptions used in the justification to support the procedure change (i.e. a license amendment was not deemed required to support the procedure change) to OP- 116.01 were non-conservative and the review of the issue in May 2010 did not adequately validate the assumptions (H.1(b)).
05000400/FIN-2010005-04	2010Q4	Green	H.4 H.5	Self-revealing	Inadequate Post Maintenance Test Procedure Results in Deenergization of the B Safety Bus and Loss of Decay Heat Removal	A self-revealing Green NCV of TS 6.8.1, Procedures, was identified for the licensees failure to develop an adequate procedure for the post maintenance test of the recently replaced main generator lockout relay (MGLR). Specifically, the licensee failed to ensure that the post maintenance testing (PMT) was within the clearance boundary that was established for the MGLR replacement. This resulted in the inadvertent deenergization of the B Safety Bus and the B Residual Heat Removal (RHR) pump, which was the only pump providing decay heat removal (DHR). As corrective action, the licensee entered AOP-25, Loss of One Emergency AC Bus, and restored DHR with the B RHR pump after approximately three minutes. The resultant increase in Reactor Coolant System temperature was approximately one degree. Additionally, the licensee plans to revise PLP-400, Post Maintenance Testing, to provide the work planner with additional guidance in the development of PMT for protective relays. The licensee entered this issue into their CAP as AR #431732. The licensees failure to develop an adequate procedure for the post maintenance test of the recently replaced MGLR was a performance deficiency. The performance deficiency was more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone, and it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, it resulted in the inadvertent deenergization of the B Safety Bus and loss of DHR. Using IMC 0609, Significance Determination Process, Phase 1 screening worksheet of the SDP, the inspectors determined that the use of Appendix G, Shutdown Operations Significance Determination Process, was necessary. Using Checklist 3 of Attachment 1 of Appendix G, the inspectors determined that this issue affected both the DHR equipment guidelines and the emergency electrical bus guidelines and therefore required a Phase 2 analysis. Using Worksheet 8 of Attachment 2 of Appendix G, the inspectors determined that recovery credit was appropriate because 1) sufficient time was available to implement these actions, 2) environmental conditions allow access where needed, 3) procedures exist, 4) training was conducted on the existing procedures under conditions similar to the scenario assumed, and 5) any equipment needed to complete these actions is available and ready for use. Using a time to boil of greater than one hour and the fact that the steam generators were not available for cooling, the result of the Phase 2 was that a Phase 3 was necessary. A regional Senior Reactor Analyst evaluated the performance deficiency under the Phase 3 protocol of the Significance Determination Process. Based upon the results of that evaluation, the performance deficiency was characterized as of very low safety significance (Green). The finding has a cross-cutting aspect of Work Coordination, as described in the Work Control component of the Human Performance cross-cutting area because the licensee did not understand the potential operational impact of the work activities or adequately account for current plant conditions (H.3(b)).
05000400/FIN-2010005-05	2010Q4	Green	H.11 H.12	Self-revealing	Failure to Follow Procedure to Properly Align the MOC Switch Contacts Associated With Breaker 1A-6 Results in Actuation of the B MDAFW Pump	A self-revealing Green NCV of TS 6.8.1, Procedures, was identified for the licensees failure to correctly implement Section D.2.10 of Engineering Change (EC) #74866R1 when aligning the Mechanism Operated Cell (MOC) switch for the A Main Feed Water Pump (MFP) breaker 1A-6. Specifically, the misalignment of the MOC resulted in the inadvertent auto actuation of the B Motor Driven Auxiliary Feed Water (MDAFW) pump. As corrective action (AR #432568), the licensee realigned MOC switch contacts under task 3 of Work Order (WO) #01658137 per the instructions of EC #74866R1. Post Modification testing verified contact continuity in both the breaker open and closed and was completed satisfactory. The failure to follow Section D.2.10 of EC #74866R1 on WO #01658137 task 1 was a performance deficiency. The performance deficiency was more than minor because it is associated with the human performance attribute of the Mitigating System cornerstone, and it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the misalignment of the MOC resulted in the inadvertent automatic start of the B MDAFW pump. Using IMC 0609, Significance Determination Process, Phase 1 screening worksheet of the SDP, this finding was determined to be very low safety significance because it was not a design or qualification deficiency confirmed to result in a loss of operability or functionality, did not represent a loss of system safety function, did not result in a loss of safety system function for a single train for greater than TS allowed outage time, did not result in a loss of safety function of one or more non-TS trains of equipment designated as risk significant for greater than 24 hours, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect of Human Error Prevention, as described in the Work Practices component of the Human Performance cross-cutting area because the licensee did not apply sufficient human error prevention tools to ensure the correct alignment of the MOC switch contacts associated with vacuum circuit breaker 1A-6 (H.4(a)).
05000400/FIN-2010005-06	2010Q4	Green	H.7	Self-revealing	Inadequate Procedural Guidance to Properly Lift/Land Leads	A self-revealing Green NCV of Technical Specification (TS) 6.8.1, Procedures, was identified for the licensees failure to establish and implement procedural requirements that would ensure the Program C relay wiring configuration in the A Sequencer remained in accordance with plant drawings following maintenance. Procedure OPS-NGGC-1303, Independent Verification, did not require the use of plant drawings to verify the As Built configuration when lifting and landing leads, which ultimately led to the deenergization of the A 6.9kV Safety bus during a surveillance test. The licensee took corrective action (AR #424668) and replaced the 86UV/SA relay, tested components within the circuit that could be affected, corrected the wiring issue and issued a memo to set expectations for utilizing plant design drawings when lifting/landing leads. The failure to establish and properly implement procedural guidance to maintain the Program C relay in the A Sequencer wired in accordance with plant drawings following maintenance on April 28, 2009, was a performance deficiency. The performance deficiency was more than minor because it affected the procedure quality attribute of the Mitigating System cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the leads being incorrectly landed would have prevented the A EDG from automatically reenergizing the A 6.9kV Bus. Using IMC 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors concluded that a Phase 2 evaluation was required because this finding represented a loss of safety function of the A 6.9kV safety bus. The inspectors performed a Phase 2 analysis using IMC 0609, Appendix A, Determining the Safety Significance of Reactor Inspection Findings for At-Power Situations and the site specific risk informed inspection notebook, it was determined that a Phase 3 analysis was required. A regional Senior Reactor Analyst performed a Phase 3 evaluation under the Significance Determination Process and concluded the finding was Green. The finding has a cross-cutting aspect of Documentation and Component Labeling, as described in the Resources component of the Human Performance cross-cutting area because the licensee did not effectively communicate expectations regarding the utilization of design drawings to aid in the proper completion of the verification sign-off form (OPS-NGGC-1303) (H.2(c)).
05000400/FIN-2010005-07	2010Q4	Green		Licensee-identified	Licensee-Identified Violation	During Mode 4 operation, TS 3.5.3 requires that one complete train of Emergency Core Cooling System (ECCS) shall be operable. During Mode 3 operation, TS 3.5.2 requires that two complete trains of ECCS shall be operable. Additionally, TS 3.0.4 prohibits transitioning into a Mode when the licensee has not met all of the limiting conditions for operation when the TS action would require a shutdown. Contrary to these requirements, between November 9, 2010 and November 10, 2010, the licensee operated in Mode 4 and transitioned to Mode 3 with both trains of ECCS inoperable. The licensee determined that the root cause of this issue was an operating procedure which incorrectly directed the operator to remove control power to both of the Residual Heat Removal Header Isolation Valves which provided suction to the Charging Safety Injection Pump. As corrective action, the licensee restored control power to the affected valves and revised the procedure. The licensee determined that this issue was reportable and will issue a Licensee Event Report which will be addressed in a future inspection report. This issue was identified in the licensees CAP as AR 432567. A regional Senior Reactor Analyst evaluated the performance deficiency under the Phase 3 protocol of the Significance Determination Process. Based upon the results of that evaluation, the performance deficiency was characterized as of very low safety significance (Green). The NRC\'s most current Probabilistic Risk Assessment model for the Harris plant was used. The surrogates for the performance deficiency were basic events RHR-MOV-CC-25 and RHR-MOV-CC-26, i.e., the piggyback motor operated valves, which were set to always be closed for the evaluation. The resulting dominant accident sequence was a Small Break Loss of Coolant Accident with operators failing to depressurize the Reactor Coolant System allowing core cooling via low pressure recirculation and high pressure recirculation failing due to the performance deficiency. The major assumptions for the evaluation included a thirty six hour exposure time and no recovery credit from the performance deficiency.
05000400/FIN-2010402-01	2010Q2	Green		NRC identified	Security
05000400/FIN-2010405-01	2010Q3			Licensee-identified	Security
05000400/FIN-2010405-02	2010Q3			Licensee-identified	Security
05000400/FIN-2011003-01	2011Q2			NRC identified	Offsite Power Supply Cables Submerged in Water	The inspectors identified the offsite power supply cables, connecting the switchyard to the startup transformers, were submerged in standing water in their underground bunkers. Additional inspection activities are needed to determine if the offsite power supply cables are suitable for operation while submerged in water. Pending the results of this additional inspection, an URI will be opened and designated as URI 05000400/2011003-01, Offsite Power Supply Cables Submerged in Water
05000400/FIN-2011003-02	2011Q2	Green	P.5 P.2(b)	NRC identified	Inadequate Procedure for Identifying Accumulated Gas in ECCS Systems	The inspectors identified a Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to establish adequate instructions to identify accumulated gas in Emergency Core Cooling Systems (ECCS). Specifically, the operations surveillance test procedure, OST- 1107, ECCS flow path and piping filled verification monthly interval Modes 1-2-3-4-5, Rev 29, could allow accumulated gases inside ECCS to be vented without being quantified and evaluated for potential adverse impacts on system operability. The licensee entered this in their corrective action program (CAP) as ARs #459683 and #459572. The corrective actions included the performance of UTs at 100% of the vented locations prior to venting the system to quantify and evaluate the effects of any gas discovered by the UTs. The inspectors determined that licensees failure to establish adequate instructions to identify accumulated gas in ECCS was a performance deficiency. The finding was more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, if left uncorrected the potential existed for an unacceptable void that could affect ECCS operability to remain undetected. The inspectors screened this finding in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined the finding was of very low safety significance (Green) since it was a deficiency determined not to have resulted in the loss of operability or functionality as determined by the review of UTs performed by the licensee through the PM program. While the PM UTs were performed at a lower frequency, the results provided reasonable assurance regarding operability of the ECCS. The inspectors determined that the finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee failed to implement operating experience from GL 2008-01 into station procedures (P.2(b)). Specifically, GL 2008-01 stated, in part, that Volumes that are close to impacting operability may require more sophisticated measurement.
05000400/FIN-2011004-01	2011Q3	Green	H.12	Self-revealing	Inadvertent Actuation of Turbine Driven Auxiliary Feedwater Pump Caused by Inadequate Procedure	A self-revealing Green NCV of Technical Specifications (TS) 6.8.1, Procedures, was identified for the licensees failure to develop an adequate post maintenance test (PMT) procedure for the replacement of a defective 6.9kV undervoltage relay (UVTXSB/1732). Specifically, the licensee failed to ensure that the PMT procedure CM-E0032 (UVTXSB/1732 relay replacement) established adequate steam isolation to the turbine driven auxiliary feedwater (TDAFW) pump to prevent an inadvertent actuation. This resulted in the TDAFW pump inadvertently starting and injecting water into the steam generators which caused an increase in reactor power to 100.2 percent for approximately one minute. As corrective actions, the licensee secured the TDAFW pump, restored reactor power to 100 percent, and replaced the failed relay. In order to return the TDAFW pump to operable, the licensee performed a surveillance test to meet the requirements of the PMT. The applicable procedures were placed on administrative hold for evaluation and revision. Additionally, an investigation was performed to determine further corrective actions. The issue was placed into the CAP as AR #472616. The licensees failure to develop an adequate PMT procedure CM-E0032 (UVTXSB/1732 relay replacement) to ensure adequate steam isolation to the TDAFW pump and prevent an inadvertent actuation was a performance deficiency. The performance deficiency was more than minor because it is associated with the human performance attribute of the Mitigating System cornerstone, and it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, it resulted in the automatic start of the TDAFW pump, water flowing to the steam generators, and a resultant increase in reactor power to 100.2 percent. Using IMC 0609, Significance Determination Process, Phase 1 screening worksheet, this finding was determined to be very low safety significance because it was not a design or qualification deficiency confirmed to result in a loss of operability or functionality, did not represent a loss of system safety function, did not result in a loss of safety system function for a single train for greater than TS allowed outage time, did not result in a loss of safety function of one or more non-TS trains of equipment designated as risk significant for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect of Human Error Prevention, as described in the Work Practices component of the Human Performance cross-cutting area, because the licensee did not apply sufficient human error prevention measures during the development and implementation of the PMT procedure (CM-E0032), to establish adequate steam isolation and prevent an inadvertent TDAFW pump actuation
05000400/FIN-2011004-02	2011Q3	Green		NRC identified	Failure to Periodically Calibrate Radiation Monitors.	The inspectors identified a Green Non-cited Violation (NCV) of 10 CFR 20.1501 for the failure to periodically calibrate radiation monitoring equipment. Specifically, in 2004 the licensee eliminated periodic calibrations for 64 radiation monitors used to evaluate the magnitude of radiation levels and quantities of radioactive material. The licensee entered the issue into their corrective action program as Action Request (AR) #477569. Planned corrective actions include re-assignment of all radiation monitors to a periodic calibration frequency and a design change to eliminate radiation monitors that are redundant or infrequently used. The inspectors determined that classifying radiation monitors as run-to-failure and thereby eliminating periodic calibrations was a performance deficiency. This finding was greater than minor because it adversely impacted the cornerstone objective to ensure the adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Although operational occurrences such as low sample line flow, loss of counts, detector high voltage, or loss of communication alarms could lead to identification of significant monitor problems, the failure to perform periodic calibrations and response checks could impair the licensees ability to reliably quantify radiation levels in the plant environs and in radioactivity released to the environment during normal and accident situations. The finding was evaluated using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process (SDP), and was determined to be of very low safety significance (Green) because the finding is not related to ALARA dose planning, did not result in an overexposure, and the ability to assess dose was not compromised due to the use of appropriate personnel dosimetry and frequent radiological surveys of RCA areas. This finding is not indicative of current licensee performance and therefore has no cross-cutting aspect.
05000400/FIN-2011008-01	2011Q2	Severity level IV		NRC identified	Failure to Report Required Information Related to MSIV Failure	The team identified a Severity Level IV violation of 10 CFR 50.73 for the licensees failure to include all required information in licensing event report (LER) 2010-002-00. The licensee submitted a supplemental LER to include all required information. The licensee entered this issue into the CAP as NCR 458636. The licensees failure to include all pertinent information in LER 2010-002-00 was a performance deficiency. This finding was considered a severity level IV violation in accordance with traditional enforcement as outlined in the NRC enforcement policy. 10 CFR Part 50.73, states in part that the LER shall contain the failure mode, mechanism, and effect of each failed component, if known. Contrary to this, the licensee failed to include specific information related to the main steam isolation valve failure in the LER. The finding was considered to be of low safety significance because it was not repetitive or willful, and was entered into the licensees corrective action program. The team determined that no cross cutting aspect was applicable to this performance deficiency because traditional enforcement violations are not screened for cross cutting attributes.
05000400/FIN-2011008-02	2011Q2	Green		NRC identified	Inadequate Control of Degraded Voltage Time Delay Settings Two Examples	The team identified a Green, NCV with two examples of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to properly control degraded voltage time delay setpoints. The licensee is evaluating changing the TS and field limits for the relays. Permanent corrective actions are still being evaluated by the licensee. The licensee entered these issues into the CAP as NCR 458376 and NCR 460601. The failure to properly analyze the degraded voltage time delay setpoints was a performance deficiency. The finding was considered more than minor because it affected with the Mitigating Systems Cornerstone attribute of Design Control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee had not analyzed whether electrical equipment needed to respond to an accident would be energized by the emergency diesel generators within the time considered in the accident analysis if a degraded voltage condition existed concurrent with an accident. In addition, there was reasonable doubt as to whether the permanently connected safetyrelated loads would remain available to respond to a LOOP following a nonaccident degraded voltage condition, for the duration of the time delay chosen for the degraded voltage relay. The finding was of very low safety significance since it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. The inspectors did not identify a cross cutting aspect for this finding because this finding was not indicative of current licensee performance.
05000400/FIN-2011008-03	2011Q2	Green		NRC identified	Failure to Maintain Environmental Qualification on Steam Generator Power Operated Relief Valves	The team identified a Green, NCV of 10 CFR 50.49 for the licensees failure to maintain its Environmental Qualification (EQ) program requirements on the Steam Generator Power Operated Relief Valves (S/G PORVs). While no immediate operability issues were identified, the licensee entered this issue into the CAP as NCR 459807. The licensee plans to properly place the components in the appropriate program. The licensees failure to maintain its EQ program requirements on the S/G PORVs was a performance deficiency. This finding was considered more than minor because it affected the Mitigating Systems cornerstone attribute of equipment performance to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, the S/G PORVs are required as per the steam line break analysis in Updated Final Safety Analyses Report (UFSAR) Chapter 15 to mitigate the radiological consequences of a steam line break by allowing the RCS to be cooled to the point where the residual heat removal (RHR) system can be placed in service within eight hours and be brought to cold shutdown within 40 hours after the accident. Removing the S/G PORVs from the EQ program reduced the reliability such that these valves would remain functional following a steam line break, which can subject them to a harsh environment. The finding was of very low safety significance because it was a qualification deficiency confirmed not to result in the loss of operability or functionality. The team determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance.
05000400/FIN-2011008-04	2011Q2	Green		NRC identified	Non-conservative Calculations for Motor Control Center Control Circuit Voltage	The team identified a Green, NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the licensees failure to perform adequate calculations for Motor Control Center (MCC) control circuit voltage. Immediate actions included testing the MCC contactors to address operability concerns. Permanent corrective actions are still being evaluated by the licensee. The licensee entered this issue into the CAP as NCR 460895. The failure to perform adequate calculations for MCC control circuit voltage was a performance deficiency. This finding was more than minor because it affected the Mitigating Systems Cornerstone attribute of Design Control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, there was reasonable doubt as to whether safety-related contactors associated with the MCCs would have adequate voltage to operate under degraded voltage conditions. The finding was of very low safety significance since this was a design deficiency confirmed not to have resulted in a loss of operability or functionality. The inspectors did not identify a cross cutting aspect for this finding because this finding was not indicative of current licensee performance.
05000400/FIN-2011008-05	2011Q2	Green		NRC identified	Failure to Control Design Limits for ESCW Flow Balancing	The team identified a Green, NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to control design limits for Essential Services Chilled Water System (ESCW) flow balancing. Immediate corrective actions included flow balance testing to address operability concerns. Permanent corrective actions are still being evaluated by the licensee. The licensee entered this issue into the CAP as NCR 458046. The failure to control design limits for ESCW System flow balancing was a performance deficiency. This finding was more than minor because it affected the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of the safety-related ventilation system to respond to initiating events to prevent undesirable consequences. Specifically, an operability limit was added to the ESCW flow balance procedure, based on information from a previous operability evaluation for an identified degraded/nonconforming condition. However, the operability limits established were not integrated into the plants design basis prior to being incorporated into the procedure and resulted in loss of margin and potentially affected the operability of the system. The finding was of very low safety significance because the finding did not result in a loss of safety function. The team determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance.
05000400/FIN-2011008-06	2011Q2	Green		NRC identified	Failure to Extend the Design Life for Molded-Case Circuit Breakers	The team identified a Green, NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, involving the licensees failure to include 79 safetyrelated MCCBs in the circuit breaker test program. Immediate corrective actions included review of breaker performance history to address operability concerns. Permanent corrective actions are still being pursued by the licensee. The licensee entered this issue into the CAP as NCR 460953. The inspectors determined that the failure to periodically test safety related MCCBs was a performance deficiency. The finding was more than minor because it affected the Mitigating Systems Cornerstone attribute of Design Control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not confirming satisfactory performance of safety-related MCCBs could lead to the inability of equipment to respond to design basis events. The finding was of very low safety significance because it was a test deficiency confirmed not to result in loss of operability or functionality. The team determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance.

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